ML20062K739

Amend 61 to License DPR-29,revising Min Critical Power Ratio Safety Limit to Apply to New Fuel Types & Making Minor Editorial Change

ML20062K739
Person / Time
Site:	Quad Cities
Issue date:	12/05/1980
From:	Ippolito T Office of Nuclear Reactor Regulation
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Shared Package
ML20062K743	List: ML20062K739 ML20064K435 ML20064K438
References
NUDOCS 8012300790
Download: ML20062K739 (51)

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{{#Wiki_filter:. 8(panouq[g UNITED STATES NUCLEAR REGULATORY COMMISSION o WASHING TON, D. C. 20655 g j 4,..... sd (]) COMMONWEALTH EDIS0N COMPANY AND IOWA-ILLIN0IS GAS AND ELECTRIC COMPANY DOCKET NO. 50-254 QUAD CITIES NUCLEAR POWER STATION UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 61 License No. DPR-29 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by the Commonwealth Edison Company (the Licensee) dated September 2,1980, as supplemented on October 3,1980, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the, Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. 2. Accordingly, the license is amended by changes to the Technical Speci fi-cations as indicated in the attachment to this license amendment, and paragraph 3.B of Facility License No. OPR-29 is hereby amended to read as follows: 8012soo go

i e B. Technical Specifications The Technical Specifications contained in Appendices A and B, /. as revised through Amendment No. 61, are hereby incorporated in the license. The licensee shall operate the facility in l accordance with the Technical Specifications'. J 3. This license amendment is effective as of the date of its issuance. FOR THE NUCL' EAR REGULATORY COMMISSION s j $*hpbdo ~ l Thomas A. Ippolito, Chief Operating Reactors Brar.ch #2 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: December 5,1980 l l l l 1

i s a A f ATTACHMENT TO LICENSE AMENDMENT NO. 61 FACILITY OPERATING LICENSE NO. DPR-29 7 DOCKET N0. 50-254 1. Remove the following pages and insert identically numbered pages: i i i 4 h, -/ +. y 3.3/4.3-3 l.0-2 3.3/4.3-4 '4-l.0-4 3.3/4.3-8 l7 - ' 1.1/ 2.1 -1 3.3/4.3-9 1.1/2.1 -2 3.3/4.3-10 i l.1/2.1-4 3.3/4.3-11 1.1/ 2.1 -5 3.4/4.4-3 s 1.1/ 2.1 -6 3.5/4.5-7 E T 1.1/2.1-7

3. 5/4. 5-9 E

1.1/2.1 -8 3.5/4.5-10 l 1.1/2.1-9 3.5/4.5-11 i 1.1/2.1-10 3.5/4.5-14 1.1/2.1-11 3.5/4.5-15 l l.2/2.2-2 3.5/4.5-18 1.2/2.2-3 l 3.1/4.1 -1 j 3.1/4.1 -3 3.1/4.1 -5 1 3.1/4.1-7 /b;' 3.2/4.2-5 3.2/4.2-6 3.2/4.2-7 3.2/4.2-8 1 3.2/4.2-14 3.2/4.2-15 2. Page 1.1/2.1-2a is added. 3. Figure 2.1-2 is deleted. l 4. Figure 3.5-1 is being replaced by 6 pages. i i 6 4 4 i / 7 - if > Ad.

I QUAD-CITIES DPR-29 TABLE OF CONTENTS (Cont'd) Page 3 9/4 9-1 3.9/4.9 AUX 1LIARY ELECTRICAL SYSTEMS 3.9/4.9-1 A. Normal and Emergency A-C Auxiliary P6wer s 3.9/4.9-2 B. Station Batteries 3.9/4.,9-2 C. Electric Power Availability 3.9/4.9-3 D. Diesel Fuel 3.9/4.9-3 E. Diesel-Generator Operability 3.9/4.9-5 3.9 Limiting Conditions for Operation Bases 3.9/4.9-6 4.9 Surveillance Requirements Bases 3.10/4.10-1 3.10/4.10 REFUELING 3.10/4.10-1 A. Refueling Interlocks 3.10/4.10-2 B. Core Monitoring 3.10/4.10-2 C. Fuel Storage Pool Water Level 3.10/4.10-2 D. Control Rod and Control Rod Drive Maintenance 3.10/4.10-3 E. Extendci Core Maintenance' 3.10/4.10-3 F. Spent Fuel Cask Handling 3.10/4.10-4 3.10 Limiting Conditions for Operation Bases 3.10/4.10-6 4.10 Surveillance Requirements Bases 3.11/4.11 HIGH ENERGY PIPING INTEGRITY 3.11/4.11-1 (Outside Contrinment) 3.11/4.11-2 \\ 3.11/4.11 Bases 3.12/4.12 FIRE PROTECTION SYSTEMS 3.12/4.12-1 A. Fire Detection Instrumentation 3.12/4.12-1 B. Fire Suppression Water System 3.12/4.12-2 C. Sprinkler Systems 3.12/4.12-3 3.12/4.12-4 D. CO2 Systems E. Fire Hose Station 3.12/4.12-4 F. Penetration Fire Barriers 3.12/4.12-4 G. Fire Pump Diesel Engine 3.12/4.12-5 3.12/4.12 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS BASES 3.12/4.12-6 5.0 DESIGN FEATURES 3,o-3 6.0 ADMINISTRATIVE CONTROLS 6.I-I 6.I Orgrnization, Review. investigation.and Audit 6.1-1 6.2 Plant Operating Procedures 6.2-1 6.3 Action to be Taken in the Event of a Reportable Occurrence in Plant Operation 6.3-1 6.4 Action to be Taken in the Event a Safety Limit is Exceeded 6.4-1 6.5 Plant Operating Records 6.5-1 6.6 Reporting Requirements 6.6-1 ( 6.7 Environmental Qualification 6.7-1 l i Amendment No.,'d, Afi, 61 iii

i i r .i ! QUAD-CITIES f.' DPR-29 l TF.CHNICAL SPF.CIFICATIONS APPF.NDIX A I' ii LIST OF FIGURES ~ k Numeer Title 2.1-1 APRM Flow Reference Scram and APRM Rod Block Settings 2.1-2 Deleted 2.1-3 APRM-Flow Bias Scram Relationship to Normal Operating Conditions Graphical Aid in the Selection of an Adequate Interval Between Tests 4.1-1 4.2-1 Test Interval vs. System Unavailability 3.4-1 Standby liquid Control Solution Requirements 3.4-2 Sodium Pentaborate Solution Temperature Requirements Maximurti Average Planar Linear Heat Generation Rate (MAPLHGR) vs. Planar Average Exposure 3.5-1 3.5-2 K,. Factor 3.6-1 Minimum Reactor Pressurization Temperature 3.12-1 Fire Detection Instruments 3.12-2 Sprinkler Systems 3.12-3 CO2 Systems 3.12-4 Fire Hose Stations Chloride Stress Corrosion Test Results at 500*F 4,6-1 Corporate Organization 6.1-1 Station Organization Chart (Two Units at flot Shutdown or Power) 6.1-2 6.1-3 Minimum Shift Crew Composition I i l V Amancknent No. A5, 61 i

QUAD-CITil'.S 5 DiH-29 9 H. Umi In: Conditions for Operation (LCO) The limiting conditions for operation specify t acceptable levcis of system performance newss;rry to anure safe stattop and ope When these conditions are met, the plant can be operated safely and abnormal situations c controlled. Umiting Ssfety System Setting (LSSS) The limiting safety system settints are settin ill not he tion which initiate the automatic protective action at a level such that the safety limits w L eacceded. The rc5 on between 'the safety limit and these sellings represents mar eptration lying below these settings. The margin has been established sd i the lastrumentation, the safety limits will never be exceeded. Iagic System Functional Test - A logie system functional test means a test a logic circuit from sensor to activated device to ensure all components are op K. Where possihte, action will go to completion; i.e., pumps will be started and valves Modes of Operation - A reactor mode switch selects the proper interlocking for the shutdown condition of the plant. Following are the modes and interlocks provided: L

1. Shutdown - In this position, a reactor scram is initiated. power to the control r and the tractor protection trip systems have been deenergized for 10 conds prio manual reset.

g

2. Refuel - In this position,i;terlocks are established so that one control rod only m i

h when flux amplifiers are set at the proper sensitivity level and the refueling crane s reactor. Also, the trips from the turbine control valves, turbine stop valves, mai valves, and condenser vacuum are bypassed. If the refueling crane is over the rea be fully inserted and none can be withdrawn.

3. Startup/ Hot Standby -In this position,the reactor protection scram stip. in vacuum and main steamline isolation valve closure, are bypassed, the low prewu isolation valve closure trip is bypassed, and the reactor protection system is e APRM neutron monitoring system trips and control rod withdrawalinterlocks in serv

4. Run In this position the reactor system preoure is at or above 850 psig, and system is energized. with APRM protection and RMB interlocks in serv Aus scram)-

M. Operable A system or component shall be considered operable when intended function in its required manner. N. Operating Operating means that a system or component is performing required manner. O. Operating Cycle Interval between the end of one refueling outage for a v.te next subsequent refueling outage for the same unit. Primary Containment Intetrity Primary containment integrity means that the drywell P. supprenion thamber are intact and all of the following conditmns are satisfied: g

1. All manual containment isolation valves on lines connecting to the reactor coolant s containment which are not required to bc open during accident conditions are closed.

1.0-2 Amendment No. 61 ~ y

QtfAI).CITIF.S DPR-29 O v Y. Shuldnun. The re.ictor is in a shutdown condition when the reactor mode switch i. in the Shutdown position and no core alterations are being performed. I. Hot Shutdown means conditions as above, with reactor coolant temperature greater than 212' F.

2. Cold Shutdown means conditions as alxwe, with reactor coolant temperature equal to or few than 212 F.

2. Simulated Autoinetic Actuation. Simulated automatic actuation rWans applying a simulated signal to the sensor to actuate the circuit in question. BB, Transition Balling. Transition boiling means the boiling regime between nucleate and I;Im boiling. Transition boiling is the regime in which bo:h nucleate and film boilmg occur intermittenity.with neither type being cumpleiety stable. CC. Critical Power Ratio (CPR). The critical power ratio is the ratio of that auembly power which causes some point in the auembly to experience transition boiling to the assembly power at the reactor condition ofinierest as calculated by application of the GEXL correlation (reference NEDO.10958). DD. Mialmum Critical Paner Ratio (MCPR).The minimum incore critical pcwer ratio corresponding to the most limiting fuel assembly in the core. EE. Surveillance Intenal. Each surveillJnce requirement shall be performed within the specified surveil-lance interval with:

a. A masimum allowahle extension not to exceed 25% of the surveillance interval.

b. A total maximum combined interval time for any 3 consecutive surveillance intervah not to euced 3.25 times the specified surveillance interval FF.

Fraction of Limiting Power Density (FLPD) - he fraction of limiting power density is the ratio of the linear heat generation rate (LHGR) existing at a given location to the design LHGR for that bundle type. GG.. Maximum Fraction of Limiting Power Density (MFLPD) - The maximum fraction of limiting power density is the highest value existing in the. core of the fraction of limiting power density (FLPD). IDI. Fraction of Rated Power (FRP) - he fraction of rated power is the ratio of core thermal power to rated thermal power of 2511 MWth. I 1.0-4 r ( Amendment No. 61

i. s QUAD.CITtrS Di& 29 1.1/2.1 FUI'.1. CLADDING INTI:GitlTY IJMITING SArl:TY SYS1DI SrrilNG sAFETYf.nllT Appliental!It)1 Applicabilii3: The sarciy limits established to preserve the fuel The limiiiaf **f) $y$'em 'r" int

• arr') to trir claddirig integrity apply in thee vatut.fes which settings of she instruments and devices sliish are Provided to paesent the fuel etadding integrity snordtor tne fuel thermal lacha ror.

esfety limits froin being esteede l. Objeeilie: 06},eti.e: The objective of the sarciy limits is to esiahlish The objective of the hmiting safety spiem settints is to deAne the level of the proce+s varutites at v hich limits b: low = hich the integrity of the fact cladding automatic proiective aci on is iniuated to prevent is preserved. the fuel cladJmg iniquiy saft y hmits inom tont sacreded. SPECil'ICATIONS O l A. Reerior Precuee > 500 rig and Core flow A. Neutron I'lus Trip Settinr.s > ler.ef Resed The limiting saftiy system trip settings shall be %e eXISterCe Of a mimiN as specinrd be'uw. critical power ratio (frPR)

1. Arnu rlus seram Trip sening inun less than 1.07 shall constitute gode) violation of the fuel clad-w y,,,,,,,,,,,,,,,,,;,,,, ;, ;,,,;,

ding integrity safety hmit. go, pos;i;on,,i,e ai.ity glu,,c,am actting. stuli 1.c as shsv n in Ist m e 2.1 1 and shall I c. B. Core Dermal Power unmis (Reettor Pre %ure 5s(45W + $5) l D s 800 psi ) orish a monimum seipoint of 120'*. for When the recetor pressure is s 800 psig or '4"'I '" N " # este Aow is less chan 10'". of rateel, the core I" shermal power shall not execed 2.W of rated where-thermal power. Settinf, in Percent of sated S C. Peeer Transient power I. The neutron flus shaft riot esceed the Wp percent of drive now re-l actam setti'ig est bbshed iin Specifica. , g, g g,,,,,,g,c,,,,, g cc,,,,,, tion 2.l.A for lon;cr then 1.5 seconds g3ew og ga 111 ton Ib/t.r. In as indicated by the proects computer. then event of opeaestuu wath a mentmass f raction of limitifv7

3. When the process compuier is out of power donesty tartxtil greater service this s1fely hmii stuli la as.

than the t-*essos of cated sumed la be esecedent er the neuuon gener ify -t t i rvi shall be moditled _. flua eseccds the strani wtiinT c.s ih. lashed liy $.*ufic twn 2.l A and a [*,7fP '} 3 eentrol ruJ stram does hot meut. S4(.65WD + SM L Enft' i 1.1/2.1-1 i Amendment No. 61 6

QUAD-CITIES DPR-29 O where Reactor Water lect (Shutdown Conditiori) D. FRP = fraction of rated whenever the reactor is in the shut-thermal power down condition with irradiated fuci (2511 MWt) in the reactor vessel, the water MFLPD = maximum fraction of level shall not be less than that limiting power dens-corresponding to 12 inches above the ity where the limit-is top of the active fuel

• when it ing power density g

seated in the core. for each bundle is the design linear

• Top of active fuel is defined to be heat generation rate 360 inches abovo vessel zero (See for that bundle.

sases 3.2). The ratio of FRP/MFLPD shall be set equal to 1.0 unless the actu-al operating value is less than 1.0 in which case the actual operating value will ba used. This adjustment may also be performed by increasing the APRM gain by the inverse ratio, MFLPD/FRP, which accomplishes the same degree of pro-tection as reducing the trip setting by FRP/MFLPD. 2 ^rax etex scr>m T :e set'i a tae-O fueling or Startup and Hot Standby Mode) When the reactor mode switch is in the Refuel or Startup Hot Standby posi-tion, the APRM scram shall be set at less than or equal to 157, of rated neutron flux.

3. IRM Flux Scram Trip Setting The IRM flux scram setting shall be set at less than or equal to 120/125 of full I

scale.

4. When the reactor mode switch is in the startup or run position, the reactor shall j

not be operated in the natural circula. tion flow mode. B. APRM Rod Block Setting The APRM rod block setting shall be as shown in Figure 2.1-1 and shall be-S s (.65Wo+ 43) 1.1/2.1-2 g l l 1 l f Amendment No. 61

t QUAD-CITIES non-29 The dofinitionc used above for the APAM scram trip apply. In the event of oper-ation with a maximum f raction limiting powcr density (MELPD) greater than the fraction of rated power (FRP), the setting shall be modified as follows: FRP._ S S (.65Wp + 43) MFLPD The definitions used above for the APRM scram trip apply. The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than 1.0, in which case the actual operating value will be used. This may also be performed by increasing the APRM gain by the inverse ratio, MFLPD/FRP, which accomplishes the same degree of pro-tection as reducing the trip setting by FRP/MFLPD. .C. Reactor low water level scram setiting shall be 144 inches above the top of the active fuel

• at normal operating condi-P tions.

D. Reactor low water level Eces initiation shall be 84 inchos (+4 inches /-0 inch) above the top of the active fuel

• at normal operating conditions.

+ f E. Turbine stop valve scram shall be s 10% valve i i closure from full open. 1 F. Turbine control valve fast closure scram shall initiate upon actuation of the fast closure sole-noid valves which trip the turbine control valves. I G. Main steamline isolation valve closure scram ~ shall be s 10% valve closure from full open. H. Main steamline low. pressure initiation of main ( steamline isolation valve closure shall be 2 850 psig.

• Top of active fuel is defined to i.

be 360 inches above vessel zero (See Bases 3 2) () 1.1/2.1-2a Amendment No. 61

P ) v OUAD-CITIES nN-29 1.1 SAFFTY LIMIT 15 ASIS The fuel cladding integrity limit is set such that no calculated fuel dwsage would occur as a recu1* of an abnormal oper.it iona l tr)nsient. Occause fuel damage is not directly cbacavabic, a step-back apps occh is used to establich a safety limit such the the mansr un critical power ratio (MCPl:) an no less th:in the fus t clarding integrity saf ety linit 0 Crit > the fuel cladding integrity safety limit representa a conbervative margin relativo to the condataons required to maintain fuel c1codang integrity. The fuel cladding is one of the physf cal hara t ers 6hich separate radioactive saateri.ils fro a the environs. The integrity of this cladding bara ter la reinted to its relative f recJcus from person etions or crackang. Although some corrosion or usc-relatcd cracking nay occur during the life-of the cladding, fazzarn proJuct migraticn from this sourec is incrementally curmlnt ive-and continuou=1y measurabic. Fuel cladd ang per-forations, however, can result from thermal staetes khich occur fro.a s cactor operatten signific ntiv above design conditions and the protectiori systen c.sfety settings. Uhtle fluson product migratien f res clyddinj perforation is ju*.t as measurable as that from uec-related cracking, the thermally couted claddia.g cer Na. ations signal e threshold beyond wh'ich still greater thermal stresses snay cause ga oss rather than a r. cts.w nt-al cladding deterioration. Therefore, the fuel cladding safety limit is defined with raargan to the condi-tions shich would produce onset of transation boaltng (MCPA of 1.0). Ther.e conditionr. represent a regniti-cent departun e frcr9 the condition intended by der.ign for planned operation. Therefore, the fuel elev'Atng integrity safety limit is tstablished such that no calculated fuel d.anage shall result I rtim an abnormal operataonal transaent. Basic of the values derived for this cafety limit for each fuel type is documented in Reference 1. A. Reactor Pressure '/ GOO peig and Core Flow > lt4 of mated Onset of transition boiling results in a decrease in heat trannter f rom the claddle.g and there fore elevated cladding teeparrture and the possacility of cladding f allus o. tiowever, the exar te nce of critical power, or bosling transition 16 not a directly observablu paremoter a n an ops rat ang r-act-or. Therefore, the margin to boilang transition is calculated frcu gelant oporatang parrmetens such as core power, core flov, feesweter temperature, a nd c ore powe r dis t s a bu t s en. The margin f at-es e?. fuel assembly is characterired by the critical power ratio (CPH), which 15 'h4 ratio of the Sa ndle T. power which would prmduce onset of transataon boiling dividad by thu actual bm.dlo po.c r. The minimurs valuc of this ratio for any bundic in the coro is tha minimum critic 1 power ratio (N pn). It is assumed that the plant operation is controlled to the ncesinal protect &ve netponts via tho instrvmonted vertables (rigure 2.1-3). The NCpR fuel cladding integrity safety lisait has suf ficient con;crvctita to esbure that in the ever.tl of an abnormal operational transient initiated from the norn.il opciating con l s t ion, mic r e tnan 99 n3 of the fuel rods an the coro are expected to avoid boiling transit ton. The n.arg s n beten f.cP;t of 1.0 (oncut of transition boiling) and the ma tety limit, is derived f rom a detailed stat tstacal g analysin considering all of the uncertaintaes in pionitoring the core opernstig state, including oncertainty in the bollang trancition cortulation (see e.g., he terence 1). Because the boalang transitjen correlation is hated on a larce quantity of full-scale dat a, there a n a very high con. fidence that operation of a fuel assembly at the condition of MCru - the fuci c1mJding integrity l safety limit would not produce boiling transation, a Itowever, is boiling transition were to occur, eladding perforation would not be expected. Cardding temperature s would increanc to appronanntely 1100 r, which is below the perforation tampuratute of the cladding amaterial. This han been verified by tests in the cenaral Electric test Reactor (CCTA). where similar f uel operated above the critical heat flux for a significant pcraod of time (3C miin-utes) without cladding perforataan. If reactor pressure should ever exceed 1400 psia during normal power operation (the limit of applicability of the boiling trancation correlation), it would be ncsumed that the f uel cladding integrity safety limit has been violated. In addition to the boiling transition livnit (MCPR) operation is const rained to a maximum L H CR s 17,5 kw/ft for 7 x 7 fuel and 13.4kv/ft for all 8x8 fuel types. This constraint is entsb11shed by specification 3.s.3. to orovide adequate safety maroin to 1% plastic strain for abnormal operating transients initiated from high power conditions. Specification 2.1.A.1 provides for equivalent safety margin for transients initiated from* lower power con-ditions by adjusting the APRM flow-biased scram setting by the ratio of FRP/MFLPD. m I \\.) 1.1/2.1-4 Amendment No. 61

QUAL) Cilll'.S O ovn-29 I Speciiication 3.5J established the LilGR maximum which cannot be exceeded under stead operation. Core Thermal Power Limit (Reactor Preuvrc<800 psla) 8. At pressures below 800 psia, the core elevation pressure drop (0 power,O flow) is gre At low powers and flows this pressure differentialis maintained in the bypass region of th the pressure drop in the bypass region is essentially all elevation head, she core pres powers and flows will always be greater than 4.56 psi. Analyses show that with a f bundle flow, bundle pressure drop is nearly independent of bundle power and has a value of 3. the bundle flow with a 4.56. psi driving head will be greater than 28 x 10'lb/hr. Full scale ATLAS data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critica flow is approximately 3.35 MWt. At 25% of rated thermal power, the peak powered bund to be operating at 3.86 times the average powered bundle in order to achieve this bundle a core thermal power limit of 25% for reactor pressures below 800 psia is conservative. C. Power Tranglent 'During transient operation the heat flux (thermal power.to. water) would lag behind the to the inherent heat transfer time constant of the fuel, which is 8 to 9 secon$. Also. the limiting h system scram settings are at values which will not allow the reactor to be oper,tted a during normal operation or during other pla,it operating situations u hoch have been analp 9 In addition, control rod scrams are such that for normal operating transients, the neutron flux tran Control rod scram times is terminated before a significant increase in surface heat flux occurs. are checked as required by. Specification 4.3.C. Exceeding a neutron flux scram setting and a failure of the control rods to reduce flux to le the scram setting within 1.5 seconds does not necessarily imply that fuel is damaged; howeve specification, a safety limit violation will be assumed any time a neutron flan scram for longer than 1.5 seconds. If the scram occurs such that the neutron flux dwell time above the limiting safety system settin than 1.7 seconds, the safety limit will not be exceeded for normal turbine or generator trips, wh the most severe normal operating transients expected.These analyses show that even if the bypass s g fails to operate, the design limit of MCPR - the fuel cladding intecrity safety l limit is not exceeded. Thus, use of a 1.5 necond limit provides The computer provided'has a sequence annunesation progrard which willindicate the s additional marcin. scrams occur, such as neutron flux, pressure.cic. This program also indicates when the scram cleared. This will provide information on how long a scram condition exists and thus provide some measure of the energy added during a transient.Thus, computer information normally will be for analyzing scrams; however, if the computer information should not be available for any scram analysis, Specification 1.1.C.2 will be relied on to determine if a safety limit has been violat During perinds when the reactor is shut down, consideration must also be given to wat requirements due to the effect of decay heat. If reactor water level should drop below fuel during this time, the ability to cool the core is reduced. This reduction in enre. cooling ca could lead to elevated claddinr. temperatures and cladding perforation.The enre will he evnted s to prevent stadding melting should the water level be reduce (to two. thuds the rose h ment of the safety hmii at 12 inches above the top of the fuel provides adequare mar;.m. 't h g be continuously monitored whenever the rniisulaiion pumps are not operating.. 360 inches above vessel

• Top of the active fuel is defined to be zero (see Bases 3.2).

1.1/ 2.1 -5 Amendment No. 61

e Qt'AD Cllll3 a. pl'u g s References " Generic Reload fuel Applications," NEDE-24011-P-A* 8 Approved revision number at time reload fuel analys'es are p'erfomed. i z, i j 1 i l 8 ..t t. 's l \\. I i.i n.i.6 i i o ..r t j ' Amt ndment No. 61 ,.f. i L f.

QllAD-CliILS /~ DPib29 V 2.11.lMITING SAll3 Y SYSll M SETTING IIASES The abnormal operational transients applicable to operation of the units have been analyred throughout the spectrum of pl.rnned oper.ruar sondstions up to the rated thermal power condition of 25 I I MWt. in. MWt is the inensed masimum steady. state power level of the units. This maximum steady. state power level w never Lnowingly be exceeded. Conservatism is mcorporated in the transient analyses in estimating the controlling factors, such as voi coethcient, control rod u ram

• orth, scram delay time, peaking factors. and axial power shapes. These factors selected conservatisely with resocci to their effeu on the applicable transient res analysis model.

la documented in Reference 1. Transient analyses are initiated at the g conditions given in this Reference. Ihe atnolute value ci sne voto reacuvny coemesent useo in tne anstyns is conservativeiy estimateo to ce snout a.o greater than the norninal maximum value expected to occur during the core Ufetime. The scram been derated to be equivalent to approximately 80".of the total scram worth of the control rodt 1he scram delay time and rate of rod insertion allowed by the analyses and conservatively set equal to the longe >t delay and slow l insertion rate acceptable by techmcal specifications. The effects of scram worth, scram delay time, and rod I rate, all conservatnelv apphed, are of greatest significance in the early portion of the negative reactivity ins The rapid ansernon of nepaine reactivity is assured by the time requirements for 5% and 20%insertron ume the rods are on9 mserted. approximately 4 dollars of negative reactivity have been inserted, which strongly 1 (i i tuens the transient and accomplishes the desired effect. The times for 50% and 90% insertion are g ven to auure V proper completion of the expested performance in the earlier portion of she transient, and to establish fully shut down steady uate condition This choice of using conservative values of controlling parameters and initiating uansients at the design power level produces rnore pessimntic anwers than would result by using expected values of control paramete analyzing M higher power levels. Ste.rdy. state operation = sihout forsed recircut.: lion will not be permitted except during startup testing. to support operatmn at various power and flow relationships has considered operation with either one or two recirculation pumps. The bases for individual trip settings are discussed in the following paragraphs. 1 For analpes of the thermal consequences of the transients, the MCPR's stated in Paragraph 3.5.K l as the limiting condition of operation bound those which are conserva-tively assumed to eiist prior to initiation of the transients. A. Neutron i lus 1 rip Settings l

1. APRM Hua Scram Trip Setting (Run Mode)

The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady. state conditions, reads in percent of rated thermal power. Because fusion chambers proside the basic input signals, the APRM system responds directly to average neutron flux. During tranuents, the instantaneous rate of heat transfer from the fuel (reactor thermal powei) is less than the instantaneous neutron flux du.: to the time constant of the fuel. Therefore. during abnormal' operational tr.insients. the thermal power of the fuel will be less than that indicated by the neutron flus at the st ram seuinp Analyses demonstrate that with a 120% scram trip seuinp. none of the abnormal operanonal tr.insienn analyzed viol.ites the fuel safety linut, and there is a substantial margin trom fuel damage. Therefore the use of flow-referenced scram trip provides even additional g ,l mar gin. I 1.1/ 2.1-7 Amendrnent No. 61 1r

T y u rw ~. - - --~ DPR-29 in the APM acram trip setting would decrease the margin present before the The APM scram trip setting was determined An increase f fuel clsJJany integrity saf ety limit, is reached. required to provide a reasonable range for man by an analysts of margins increase the frequency of spurious scrams, keducing this operating margin would operation. which have an adverse ef fect on reactor catety becaum of the resulting ther.. I stresses. it provtdeu adcauste margan ror the the APm scram trip setttng was selected tecauno allows operating margin that reduces the possibil-

Thus, fuel cladJing anteqrsty safety Itmit yet ity of unnecessary scrams.

The scram trip settir.g must be adjusted to ensure that the IJ1GR transient peak is not and I increased f or any combination of maximum f raction of limiting power density (MFLPD) The scram sotting is adjusted in accordance with the formula reactor core thermal power. fraction of rated power (FRP). in Specification 2.1.A.1, when the MFLPD is greater than the by the reciprocal The adjustment may be accomplished by increasing the APRM qainThis provide of FitP/MPLPD. the trip setting by FRP/MFLPD by raising the initial APRM readings closer to the trip settings such that a scram would be received at in a transient as if the trip settings had been re-the same point duced by FRP _ MFLPD* 2. APM Flux Scram Trip Setting (Refuel or Startup/ Hot Standby Mode) For operation in the Startup mode while the reactor is at low pressure, the APRM scram settir of 15% of ratcd power provides adequate thermal margin between the setpoint and the safety limit, 25*i of rated. The margin is adequate to accomodate anticioated maneuvers assoctsted with power plant startup. Ef fects of increastnq pressure at z:ro or low 40id c0nte.t 3re siinor, col.1 anter fr:- JNrm evn11abla Juring atartup La not much co1Nr ma tWL ulcendy in ta: c.c e' te tent s.,:e small, and control rod patterns are constrained to be system, t em s: t a w r.: uniform by operating procedures La:md uo by the rod worth minimizer. Of all possible source of reactivity input, uniform control rod withdrawal is the most probable cause of significan: i } power rine. Because the flux distribution associated with uniform rod withdrawals does not i involve high local peaks, and because several rods must be moved to change power by a signif; cant percentage of rated power, the rate of power rise is very clow. Generally, the heat f it la in near equilibriun with the fission rate. In an assumed uniform rod withdenwal approach r to the scram level, the rate of power rise is no more than 5% of rated power per minute, and the APRM system would be more than adequate to assure a scram before the power could exceed the saf ety limit. The 15% AFRM scram remains active until the mode switch is placed in the Aun position. This switch occurs when reactor pressure is greater than 850 psig. [ 3. IM Flux scram Trip setting i } The IM system consists of eight chambers, four in each of the reactor protection system log: channela. The I M is a 5-decade instrument which covers the range of power level between thi j covered by the SRM and the APRM. The 5 decades are broken down into 10 ranges, each being one-half a decade in size. I The IRM scram trip setting of 120 divisions is active in each range of the IRM. For cample. { if the instrunent were on Range 1, the scram setting would be 120 divisions for that ranger i likewise, if the instrument were on Range 5, the scram would be 120 divtsions on that range. j Thus, as the I M is ranqed up to accommodate the increase in power level, the scram trip set. l ting is also ranged up. The mont significant cources of reactivity chango during the power increase are due to contre rod watt.!rawl. In order to ensure that the IRM provides adequate protection against the single rod withdrawal error, a range of rod withdrawal accidents was analyzed. Tnis analyst. included starting the accident at various power levels. The most severe case involves an initial condition in which the reactor is just suberitical and the IRM system is not yet on scale. Additional conservatism wa9 taken in this analysic by esruming that the IRM chsenelC10Str.t

v. h n m ed.

The results of tht3 analvsts show that the reactor is scrc.-m 4e the wl h lr wn r e t i and p. m p.mcr L amtt. a to IS of rated power, than mainta'ining MCPR above the fuel cladding inteqrsty nataty Itmtt. Dased on the above analysis, the IRM providos protection against local control ro.1 withdrawal errors and continuous withdrawal of control rodo in sequence an provides backup protection for the APRM. 1.1/2.1-8 Amendment No. 61 w ~'

QUAD-CITIES =:+ . non-29 g 4 V 3 APM Rod alock Trip Setting rectreulatien flow Reactor power level may be varied by moving control rods or by varying the The APRM system provides a control rod block to prevent gross rod withdrawal at constant rate. recirculation flow rate to protect against grossly exceeding the MCPR l'uel Cladding Integrity Safety Lusit. This rod block trip setting, which is automatically varied with recirculation loop f1w rate, prevents an increase in the reactor power level to excessive values due to The flow variable trip setting provides substantial margin from fuel control rod withdrawal. damage, assuming a steady-state operation at the trip setting, over the entare recirculation flow range. The margin to the safety limit ir. creases as the flow decreases for the s peci fied therefore the worst-case MCPR which could occur during trip setting versus flow relationshipt steady-state operation is at 108% of rated therrial power because of the APRM rod block trip The actual power distribution in the core is established by specified control rod setting. sequences and is n'onitored continuously by the incore LPP.M system. As with APM scram trip fraction of limit-i setting, the APRM rod block trip setting is adjusted downward if the maxinura ing power density exceeds the fraction of rated power, thus preserving the APRM rod block safety margin. C. kaactor Low water Level scram The reactor low water level scram is set at a point which will assure that the water level used is maintained. The scram setpoint is based on normal ooerat-in the bases for the safety limit ing temperature and pressure conditions because the level instroentation is density compensated. D. Reactor Inw Low water Level ECCS Initiation Trip Point suf ficient cooling to the core The emergency core cooling subsystems are designed to provide to dissipate the energy associated withthe loss-of-coolant accident and to licit fuel cladding temperature to well below the cladding melting temperature to assure that ecre geometry remasns To accomplish their intact and to limit any cladding swtal-water reaction to less than IL intended function, the capacity of each emergency core cooling system component was estaclished =; 'j based on the reactor low water level scram setpoint. To lower the setpoint of the low water level scram would increase the capacity recuirement for each of the ECCS components. Thus, the low enough to pernit margin for operation, yet will reactor vessel low water level scram was set not be set lower because of ECCS espacity requirements. the above criteria was dependent on three previously The design of the ECCS components to meet set parameters: the maximum break size, the low water level scram setpoint, and the ICCS initiation setpoint. To lower the setpoint for initiation of the ECCS could lead to a loss of effective core cooling. To raise the ECCS initiation setpoint would be in a safe direction, but it would reduce the margin established to prevent actuation of the ECCS during nor:nal operation or during normally expected transients. E. hurbineStopvalvescram ' The turbine stop valve closure scram trip enticipates the pressure, neutron flux, and he at flux i increase that could result from rapid closure of the turbine stop valvas. With a acram trip setting of 10% of valve closure from full open, the resultant ancreann in surface heat flux is limited such that MCPR remains above the MCPP. fuel cladding integrity safety limit even during g the worst-case transient that assumes the turbine bypass is closed. F. Turbine Control valve rest closure scram The turbine ?ontrol valve fast closure scram is provided to anticipate the rapid increase in pressure and neutron flux resulting frcus f ast closure of the turbine control valves due to a load rejection and subseauent isilure of the bypass, i.e., it prevents MCPR from becoming less For the Ic'ad than the MCPR fuel cladding integrity safety linst for this transtent. the peak heat flux rejection without bypass transient from 100% power,15% which provides wide (and therefore LHGR) increases on the order of margin to the value corresponding to 1% plastic strain of the cladding. 1.1/2.1-9 Amendment No. 61

e e o QUAD-CillF.S (q DPR-29 G. Reseter Coolant few Preware Initiates Main Steam Isolation Volve Closure The low. pressure isolation at 850 psig was provided to oscurs in the Run rnode when the main sacarnline isolation valves are closed to pro shutdown so that operation at pressurrs lower than th ensafe condition. H. Main Steamline Isolation su Valie Clusure Scram The low. pressure isolation of the main steamlines at 850 ps.';; was prov f rapid reactor depressurizat:on and the resulting rapid cooldown of the v i l re closed to the scram feature in the Run mode which occurs when the main steamlin provide for reactor shutdown so that high power operatton at low reac providing protection for the fuel cladding integrity safety limit. Operation of t i lower than i SO psig requires that the reactor mode switch be in the Startup position, w of the fuel claddmg integrity safety limit is provided by the IRM and APRM high neu t Thus. the combinatson of main steamline low. pressure isolation and isolation Run mode assures the availability of neutron flux scram protection over the en of the fuel cladding integrity safety limit. In addition. the isolation valve closure anticipates the pressure and flux transients which occur during normal o no increase in neutron l closure. With the scrams set at 10% valve closure in the Run moda, there is flux. O Turhine EHC Control Fluid law Preuure Scram The turbine EHC control system operates using high-pressure oll. There are system where a lost of oil pressure could result in a fast closure of the tu i closure of the turbine control valves is not protected by t control valve t'ast closure. the core would be protected hich senses ailure of r turbine control valves, a scram has been added to the reactor protection system w control oil prenure to the turbine control system. This i f 900 psig is set to that resulting from the turbine control valve fast closure scram. The scram setpoint o high enough to provide the necessary anticipatory function and low eno ill spurious scramt Normal operating pressure for this sys begin>. Condenser Low Vacuum Scram Loss of condenser vacuum occurs when the condenser can no longer han J. h condenser vacuum initiates a closure of the turbine stop valves and turbine b eliminates the heat input to the condenser. Closure of the turbine stop and bypass f li i f transient, neutron flux rise, and an increase in surface heat flux.To prevent the cladd being cacceded if this occurs. a reactor scram occur j being eseceded in the event of a turbine trip transient with bypass closure. The condemer low vacuum nram is anticipatory to the s 9 h mode at 2.1.mch lip. vacuum stop valve closure occurs at 20-inch Hg vacuum Hg vacuum. I 1.1/2.1-10 Amendment No. 61

QllAD. CITIES DPR-29 References " Generic Roload Fuel Application," 1TED -240ll-P-A* 1.

• Approved revision number at time reload analyses are performed i

e t ~i r, L ] 1.1/2.1-11 l l i .. t 8 '.rI i f t Anendment No. 61 ,3 tl 4 l

4 4 a P'sGuz., 2 1 g bag been 002egeq 0 f 67 l

{ O QUAL)411II.S nPn-29 l.2 $MT.TY IJhllT IIASI'S The reastor wolant systent insegrity is an important bstrirr in the presenison of uncontroitcJ release of feuion products. liis ewnisal slut the inieprity ofihis sysicm be protested by cstah!nhinp a pressure Innit to be observed for all operating conditions and whenner there is irradiated fuel in the reactor veswl. He precure safety hmit of 1325 pig as meeureJ by the vessel stcain space pressure indicator is equivalent to 1375 pig at the lome elev.nion of the reastor cnotant system lhe 1175 pip. value is derned from the doign pressuto of the seative incuurs vence and cnolant system pipint lhe 3DPecusc deugn gueuures ate 1250 pig at$75'Itand 18 75 per ai St>0- I.1he preuvre safety linut was chosen as.he lower of the prenure transients e permitted by the appbeable deugn todes' ASME 11 oiler and hessure Veuct Code Section lit for the preuure so.el. and USA 511111.8 Code for the reactor toolant s>siem piping The ASMt. Iloiter and Prtuure Venel Code perinits pressure transients up to In'.oser design pressure (110". x 1250 - 1375 psig), and the 15A51 Code perinits pressure tranwenis up to 2tc.over the doign pressure (120", a 18 75 = 1410 psir) 1he sarcty limit pressure of 1375 psig is referenced to the lowest tievation of the primary coulJnt s} stem. Evaluation methodology to assure that thic safety limit pressure is not execeded for any reload is docunmnted in Mcforence 1. The design basis for the re.xior preuvre vevsI makes evident the subuantial maipn of prointien against failure at the safety pressure hmat of 1375 pug lhe vessel has becn deurned for a general niembrane streu no treater than 26.700 psi at an internal preuure of 1250 psig, this is a f.ictor of 13 below th >ield tiren;th of 40.100 psi at $75' F. At the prcuure limit of i 375 psig.the general membranc sims will only be 29.400 pi. still safely below the yield strength. O The selationships of stress levels to yield strength are comparable for the primary system piping and provide a V similar margin of pruicetion at the establis!.ed safety presme lunit. The normal operating prenuie of the scactos cuolant system is 1000 psi:. l~ur the suit.n tile orloss of elecuicalluad transients, the turbuw trip scra n on generstne load rejection suam tof. ether with the turbi.c bypass system hmits the preneure to approximately 1100 pm (Refeicnces2,3 and4). In addition. pressure reber wahes have been piovided tu seduce the probabddy of the safety ahes opeistmg us the event that the tuibene bypan shou!J fail Faiatly. the urcty valves are sired to keep the teactos coolant systcm preeui. Letow 1375 psig with no credit tal.en for rehef valves dunns the postulated f ull closute of all MSIVs wahout ducs: 0 41 % positson switch) scram. Creda as taken for the neutron flus actam.howevct. l The inducci flux scram and safety valve actuation. provk!c adequate margui i below the p(ak a!!om abic veurl pressure of 1375 psg. Resetor pressure is continuously monitored in the control room during operation on a 1500 pi full scale pressure recorder. Refesences 1. " Generic Reload Fuel Application", NCDU-240ll-P-A* 2. SAR, Section 11.22 3. Quad cities 1 Nuclear Power Station firrit reload license submittal, Section b.2.4.2, February 1974 4 GE Topical Report NEDO-20693, General Electric Dolling Water Roactor No. I licencing submittal for Quad Citics Nuclear Power Station Unit 2, December 1974. Approved revision nuinber at timo relodd analysco are performed. pb 1.2/2.2-2 Amendment No. 61

QUAD CITIES DPR-29 2.2 LIMITING SAFETY SYSTEM SETTING lsASES In compliance with Section III of the ASME Code, the safety' valves must be act to open at no h1(',her than 103% of design pressure, and they must limit the reactor pressure to no more thari ))O% of dec16n pressure. Both the high neutrcn flur. scram and safety valve actuation are required to prevent cverpressurizing the reactor pre.ssure vessel and thus exceeding the pressure refety limit. The pressure scram is available as backup prote:tf on t3 the high tiux scram. Analyses are performed as described in the

• Generic Reload Fuel Application," NEDE-240ll-P-A (approved revision number at time reload analyses are performed) for cach reload to assure

~ hat the pressure safety limit is not exceeded. If the high-flux scra.t t I were to f ail, a high-pressure scram would occur at 1060 psig. i h 1 I 9 9 e 'e g t l t t 4 i [ 1.P/2.2-3 i i l Jt i Amendmet go., 6 ,,4 ; i 4 j

i l.. .1. t ,an :

s. '

E .s.. O QtfAll.Clill{S DPR-29 C3 3.1/4.1 ItEACTOft PitOTECTION SYSTEM tJM1IING CON 111TIONS FOR Orr.IIATION SURYl ll.l.ANCE ItFQlfillD11'.NTS Appfientdht): Applicaleility: App!!c, to she instrurnentat:on and.istotiated d.. Apphes to the surveill.snce of the instrumentation and a sociated dCuces w hiGh initiJte reaClor savs shich initiase a reJrtar Wram. scram. Objecthe: Oblathe: T.e awre the operebility of the reacios prowtion To specify the type and frequency aisarv:illante to .y.orm. . he applied in the protection inurumentation. ~ lr r SPECIFICATIONS A. Instrumentation systems shall be functior. ally 4. The netprints. ndnimum number of trip sys. sems, and minimum number of instrument tested and calibrated as indicated in Tahles channels that must be operable for each pnsi. 41 1 and 4.12 respntively, i i 4 M. - tion of the rcJLlor mode switch shall he as B. Daily dosing teactor powcr operation. the core V given in Tables 3.1 1 through J.l.4. The syucm 8 responic times from the opening of the sensor pow et distribution thatt iv checked for maximum contact up to and italuding the crening of the fraction of limiting power dens-g. erip actuator contacts shall not exceed 'i0 ity (MFLPD) and compared with the 8 II '* d-fracti n of rated powcr (FRP) l B* If, during operation, the maximum when operating above 2 5 rated 4 i fraction of limiting power dens-therrnal power. 'I ~ ity exceeds the fraction of rated power when operating above 25% C'. When. is ilete mined that a channel is failed s' it rated thermal power, either: in the unsafe cenditwn and Column I of sa. l bles 3.1 1 through 3.13 cannnt be rnet, that

1. the APRM scram and rod trip system must be put in the trapped enndition block settings chall be immediatcly. Allother RPS ch.:nr.ch thai mon.

lt 'I reduced to the values stor the same verrable shall c: fantt:..aai.y i] given by the cauations teued witidn A h..u:s Theirm v.st. i wi:h t:E in Specification.t 2.1..;.1 fai;cd t h innel may be untripp':J ibr a period of j '* and 2.1.D. Thi:: r.My also time not to eseced I luiur in conduct this 'be aCddmplished by testirg. As long as the trip system with the failed channel sontains at least one operable increasing the APRM gain as described channel monitoring that ume variabte. ihai "iP 5)uem inay he platcJ in the untrip;ed therein. position for short periods of time to allnw functional icstint of all RPS inutument chan. nels as specified by Tatte.l.l.1.The trip system may be in the untripped position for no more than S hours pet functior,al seu peric.it f.it thin '* " "I' 2. the pow.r di..tr lhution shall be chan9.ti : ucit that the. ma:timu:n fraction of limiting power den:sity nu longer exct:cdu the fraction of rated pos:cr. i 3.1/4.1-1 Amendment No. 6_1

QUAD-CITIES DPR-29 gallons. As indicated above, there is sufncient volume in the piping to accomm u/ . hile sufhcient of the scram times or amount of insertion of the control rods. This function s ld he volume ren: sins to accommodate the discharged water and precludes the situation in which a sc required but not be able to perform its function adequately. Imst of condenser vacuum occurs when the condenser can no longer handle heat inp h initiates a closure of the turb.ne stop valves and turbine bypass valves. which climina aux rise, and an condenser. Closure of the turbine stop and bypass valves cause d t t prevent onuts on turbine stop valve closure. The turbine stop valve closure scram function alone is a i h bypass closure. the cladding safety hmit from bemg exneded in the event of a turbine trip 23 inches Hg scram before the stop valves are ctnsed, thus the resulting transient is less severe. Scram 7 inches Hg vacuum. vacuum, stop valve closure occurs at 20 inches Hg vacuum. and bypus closure at High radiation levels in the main steamline tunnel above that due to the n times normal are an indication ofleaking fuel. A scram is initiated whenev i i ns is prevented excessive turbine contamination. Discharge of excessive amounts of radioactivity to the s te env ro limit by the air ejector olT. gas monitors, which cause an isolation of the main co speci6cd in Speci6 cation 3.8 is exceeded. The main steamline isolation valve closure scram is ses to scram By full open.This scram anticipates the pressure and flux transient which would o scramming at this setting. the resultant transient is insi nincant. 6 it to the A scattor mode switch is provided which actuates or hypasses the various scram functions a i h particular plant operating status (reference SAR Section 7 [") i and to allow repairs valve closure scram are bypassed.This bypass has been provided for flexibility dur ng startupided again i i to be made to the turbine condenser. While this bypes is in effect. protect on s prov i his mode. increases by the high. pressure scram and APRM 15% scram, respectively, which are etrecti If the reactor were brought to a hot standby condition fo i can result in an unreviewed radiological release. The manual scram function is active in all modes. thus providing for a manual me rods during all m, des of reactor operation. The IRM system provides protection against excessive power levels and sh intermediate power ranges (reference SAR Sections 7. i l d Startup/Ilot Standby rnodet in (reference SAR Section 7.43.2) Thus the IRM is required in the Refue an bases for Speci6 cation additinn, protection is provided in this range by the APRM 15% scram as discussed in the 2.1. In the power range,the APRM system provides required protection (refe M *5 IRM system is not required in the Run mode. the APRM's cover only the int provule adequate coverare in the st.artup and intermedi.ite range. l The high. reactor preuure, high.drywell pressure, reactor Inw water level, an I i scrams are required for the Startup/Ilot Standby and Run modes of plant operat on. to be operational for these modes of reactor operation. The tuihine condenser low. vacuum scrain is required only during power operation an up the unit fm) U 3.1/ 4.1 -3 f Amendment No. 61 j

QtlAl)-CITIES. nm-29 O 4.1 SURVEll.!ANCE REQUIREMENTS BASI'.S The minimum functional testing frequency used in this specification is based on a reliability analysis A. tising the concepts developed in Reretence 1.This concept was specifically adapted to the one taken twice logie of the reactor protection system. The analysis shows that the sensors are primar responsible for the reliability of the reactor protection system. This analysis makes use of

• rate experience at conventional and nuclear power plants in a reliability model for the system failure
• is denned as one which negates channel operability and which. due to its nature. is revealed on when the channel is functionally tested or attempts in respond to a real signal. Failures such as blown fases. ruptured bourdon tubes. faulted amplifiers. faulted cables, ese., which result in ' upscale

'downscale' readings on the reactor instrumentation are ' safe' and will be casily recognized by the operators during o,~ ration because they are revealed by an alarm or a scram. The channels listed in Table 4.1-1 and 4.12 are divided into three groups respecting functional test ^ These are:

1. on off sensors that provide a scram trip function (Group 't ):

2. analog devices coupled with bistable trips that provide a scram function (Group 2); and

3. devices which serve a useful function only during some restricted mode of operation. such as Startup/ Hot Standby. Refuel, or Shuidown. or for which the only practical test is one that c performed at shutdown (Group 31 The sensors that male up Group I are specifically selected from among the whole family ofindu on off sensors that have earned an v.cellent reputation for reliable operation. Actual history on this of sensors operating in nbclear power plants shows four failures in 472 sensor years. or a

-n, 0.97 x 10*/hr. During design, a gnal of 0.99999 probability of success (at the 50% confidence leve V adopted to assure that a hal.inced and adequate de. sign is achieved.The probability of a function of the sensor failure rate and the test interval. A 3 month test interval was planned for Group I sensors.This is in keepmg with good operating practiec and satisfies the design go bgic configuration utilized in the reactor protection system. To satisfy the long term objective of maintaining an adequate level of safety throughout the pla lifetime, a minimum goal of 0.9999 at the 95% confidence level is proposed. With the one-out of two .saken twice logic, this requires that cach sensor have an svailability of 0.993 at the 95% confid This level of availability may be maintained by adjusting the test interval as a function of the obse ] failure history (Reference 1).To facilitate the implementation of this technique. Figure 4.1 1 to indicate en appropriate trend in test interval. The procedure is as follows:

1. Like sensors are pooled into one group for the purpose of data acquisition.

2. The factor M is the exposure hours and is equal to the number of sensors in a group. n, times elapsed time T(M = nT).

3. The accumulated number of unsafe failures is plotted as an ordinate gaian M es an absciss Figure 4.1 l.

d. Afier a trend is established, the appropriate monthly test interval to satisfy the goal will be the t interval to the left of the plotted points.

5. A test interval of I month will be used initially until a trend is established.

Oroup 2 devices utilite an analog sensor fo!! awed by an ampliner and a histable trip circuit.Th and amplifier are active components, and a failure is almost always accompanied by a indication of the sourte of troubte. In the event of failure, repair or substitution can start immedia h An 'as is' failure is one that

• sticks' midscale and is not capable of going either up or down in respon 3.1/4.15

{ Amendment No. 61

w. QUAD.CITIF.S - DPR-29 1 O t switches, hence calibration is not upplicable; ie the switch is either on or off. Itased on the above, no calibtstion h required for these instru. ment titannels. l shall be cheded orne per day to determine if the APRM scram requires adjustment. This may The MFLPD B. normally he done by checking the LPRM readings. TIP tisces, or process computer ca a small number of control rods are moved daily, thus the peaking factors are not expected to chan l MFLPD is adequate. significantly and a daily check of the References I. M. Jacobs.' Reliability of Engineered Safety Features as a Function ofTesting Freque l. Vol. 9, No. 4. pp. 'i'J 312. July August 1968. j. l O i I 0 3.1/4.1-7 l I i Amendment No. 61 l 1_

?. I QUAD-CITIES DPR-29 3.2 LIMITING CONDITIONS FOR OPERATION BASES In addition to reactor protection instrumentation which mitiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the comequences of accidents which are beyond the operator's ability to control. or terminates operator errors before they result in serious consequences. This set of specifications provides the hmiting condilsons of operation for the pnmary system isolation function, initiation of the emergency core coohng system, control rod blod. and standby gas treatment systemt The objectives of the specifications. ire (I) to assure the eflectiveness of the protective instrumentation when required by preserving its capahihty to tolerate a single failure of any component of such systems even during periods w hen portions of such systems are out of service for matnien.mcc. and (2) to presershe the trip settings required to assure adequate performance. When necessary, one channel may be made inoperable for briefintervals to ccnduct required functional tests and calibrations. Some of the settings on the instrumentation that mitiates or controls core and contamment cochng have solerances explicitly stated where the high and low values are both critical and may have a substantial erlect on safety,it should be noted that the setpoints of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent ir. advertent actuation of the safety system invohed and exposure to abnormal situations. Isolation valves are installed in those lines that penetrate the primary containment and must he isolated durinc a lou of coolant accident so that the radiation dose limits are not exceeded during an accident condition. Actualmn of these valves is initiated by the protective instrumentation which semes the conditions for which isolatn n is required (this instrumentation is show n in Table 3.2 1 ). Such Instrumentation must he available whenever primary umtainment integrity is required. The objective is to isolate the primary containment so that the guidelines of Ill CFR 100 are not exceeded during an accident. The instrumentation which initiates primary system isolanon is connected in a dual bus arrangement. Thus the discuwion given in the bases for Specification 3.1 is apphcable here. 'Ihe low-reactor water level instrumentation is set to trip at >B inches on the level instrument (top of active fuel is defined to be 360 inches above vensel zero) arxi af tor allowinq for the full power prese.ure drop l across the steam dryer the low level trip is at 504 inchen atove vascel rcro, or 144 inches atxwe top of activo fuel. Itetrofit fixR fuel han an active fuel length 1.24 inrhes lorvler than earlier fuct slesigns, howver, premnt trip setpoints were used in the imA analysis.* 'this trip initiatec clonure of ('roup 2 anri 1 terimary contain-l ment isolation valven but does not trip the rM'irculation rnp". (reference T.AR, Section 7.7.7). Ibr a trip setting of 504 irrhen alove vessel rero anr1 a 60-ceconi valve clocure tim, the valven will te clnsnt tefore perforation of the clarkliry occurn even for the ruximum break. "he cett inq is, ther efore, arin gite. The low. low reactor levelinstrumentation is set to trip when reactor water levelis 444 inches above vessel ero (with top of active fuel defined as 360 inchen above vessel tero, 59" is M inches above the top of active fuel). This trip initiates closure of Group 1 primary containment isolation valves (reference SAR Section 7.7.2.2) and also activates the ECC subsystems, starts the emergency diesel generator, and trips the recirculation pumps. This trip setting level was chosen to be high enougn to prevent spurious operation but low enough to initiate ECCS operation and primary system isolation so that no meitine of the fuel cladding will occur and so that postaccident cooling can be accomplished and the cuidelines of 10 CFR 100 will not be exceeded. For the com-plete circumferential break of a 28 inch recirculation line and with the trip setting given above ECCS uutution and primary system isolation are initiated and in time so meet the above criteria. .The instrumentation also covers the full spectrum of breaks and meets the above criteria. e t

• Loss of coolant accident analysis for Drc= den Unit 2/3 & Quad Cities Units 1/2,

', NEDO-21allaCA. April,1979 l i' t 'i.[ l 3.2 / 4.2-5 j p Amendment No. 61 ,',l i y 1

QUAD-CITIF.S DPR-29 Venturi tubes are provided in the niain steamline as a means of measuring i i team flow, r mm inventory from the sewel during a steamhne break accident. In addition to mon to fthe instruinentatinn i> prnvided which

• uses a trip of Group i isolation o

t h t se d steam flow,in conjunction !nsis umemation is to detect a creak ' serident. main ueamline break outude the drywell. this trip seuiny of 120% of rate i loss such that fuel is not with the flow limiters and main sicamline vahe closure, limits the mass nventory h i s is weli below uncovered. fuel temperatures remain less than 1500* F, and release of radioactiv 10 CFR 100 6uidelines (refelente SAR Sections 14.2.3.9 and 14.2.3.10). his.trea Temperature monitoring instrumentation is prov i l s lis ble of eovering the entire setting of 200' F is low enough to detect fraks of the order of 5 to 10 gpm; thus i i discuued above..and for specitum of breaks. For large breaks,it is a backup to high h idelines of in Cl R 100 l failure. This are escreded. High radiation monitors in the main steamline tunnel have been provid i ident. With the instrumentation causes closure of Group i valves, the only valves required to clo l l ure, fission product established setting of 7 times normal bacLground and ma f e SAR Section i2.2.1.71 i f Pressure instrumentation is provided which trips when main steamline I this instrumentatinn results in closure of Group I isolation valm. n t eide protection against a pressu this trip function is hypaoed This function is provided primarily to prov i at 850 psig, insentory malfunction which would cause the control and/or bypas F; thus. there (reference SAR Section are no fission products available for release other than those in the reactor water O i 1.2.3 ). break in their The RCIC and the HPCI high flow and temperature instrumentation are provide CI isolation valves respective piping. Tripping of this instrumentation r l 2H9 b and required to be operable or m a inpped condition to mee Unlike is within limits. The instrumentation w hich initiates ECCS action is arranged in a one ou the reactor scram circuits, how:ver. there is one trip system associate f h f t that redundantcnte systems in the reactor protection system.The single failure criteria are met b swhng functions are provided, e.g., sprays and automatic blowdown an i i i lared inoper.ible. specification requires that if a trip system becomes ino l dthe h trettisenew of the s3 stem out ofwervice specifications of Specification 3.5 govern. This specincation pre >erses t e with re>pect to the single fadute criteria even during periods when maintenanc R does not The control rnd block functions are provided to prevent excessive contro go below the MCPR Fuel Claddina Integrity Safety Limit.T four SRM's will result in a rod bloel. The minimum instrument chJHnel r h el resguirements instrumentation to assure that the single failure craters.:. ire met.The minimum imtrum for the RBM may be reduced by one for a short period of time to allow for m This time period is only--3% of the operating time in a month and doe preventing an inadvertent control rod withdrawal. 3.2/4.24 p Arnendrnent No. 61 J

4 QUAD-C1' TIES DPR-29 Oh The APM rod block function in flow binced and prevents a cignifiennt reduction in 1 CPR, ecpecially during operation nt reduced flow. The APRM provides gross core protesetion, i.e., limitri the groos of control rods in the normal withdrawal sequence. In the refuel and startup/htt standby ruodes, the APRM rod block function is not at 127. of rated power. This control rod block provides the same type of protection in the Refuel and Startup/ Hot Standby modes as the APRM flow-binned rod block deoc in the Run modo, i.e., preventa control rod withdrawal before a sc' ram is reached. I The RBM rod block functions providen local protection of the core, i.e., the prevention of transition boiling in a local region of the core for a singic rod withdrawal error from a limiting control rod pattern. The trip point is flow biased. The worst-cace single control rod withdrawal error is analyzed for each reload to assure that, with the specific trip settings, rod withdrawal is blocked before the MCPR reaches the fuel cladding integrity safety limit. Below 30% power, the worst-case withdrawal of a single control rod with-out rod block action will not violate tho fuol cladding interir i t y safety limit. Thus the RDM rod block function is not required below this power level. The IRM block function provides local as well as gross core protection. The scaling arrangement is such that the trip setting is less than a factor of 10 above the indicated level. Analysis of the worst-case accident results in rod block action before MCPR approaches the MCPR fuel cladding integrity nafety limit. . O i A dow nscale indi.aiinn on an APRM or IK%f n.m andic.in-m the instrument has faded or is r.ot semitive enaugh. in either case the instrument u di not respond to chantes in control rod motion. and the control rod motion is thus i I prevented The dos nscale trips are set 4: 3/125 of full seale. I The SRM rod block with e 103 CPSand the deiccior not tully inserted assures that she SRM's are not withdiawn from the cose prior la commentmg rod with.trawal foi stasiup %e stram dncharge vulume high m.qct level rod l block provides annuncunon for ope ator action 1he alarm seigvint has been scluted to juovide adnguste time l to allow determination of the cause of level increase and coiretuvc action prior to automanc scrarn initiation. l. For efrective emergency core conhng for small pipe breaks. the IIPCI sysicm must function. since reacta preuure does not decrease rapdly enough so allow enher core spray or I.PCI to operc = in tunc.1he auiomatic presiurc reher function is prowJed as a b.nhp in the llPCI ni ihs s vent the llPCI does not opera.c 1he airangement of { the tripping runiacn n suite as en pim nic this funtiam m hen necen.uy and mmimne spunous opi rahon 1he inp I settings gncn in the specifn.ation are ai cquate in awuse the above cruena are met (rcicitm e S AR Nsamn f. 2 f. 3 ) t 1he speorication psewrses the cucsliveness of the sysicm ifuting periods of maintenante. intiny, or tabbration and also mmirnires the risk ofin.idversent operation. e c.. only one imtrument channel out of wavire. Two air ejutor cllt.n momion.nc pnwided and. m hen their trip point is trached, sau c an isolation of the air I ejertot ofbras hne botanon n unnatrsl w hen both imnumen% reat h their high inp point or one h.is an upscale trip and the other a dowmsale ing 1 bcse is a 15 nnnine ststiy I.ctoic the air ejretor ofbras ivdatnin s afvc is chned. This delay is accoumed for by the Emmune lu.htup time of slic otr. gas before it n rcle.ned to ihc chimney. Iloth lastruments are requised for trip, but the mstiernents are 50 designed that any imsrument failure gives a downscale trep 1he top scinnr> of the in trunnnis are set so that the chinmey reic.nc sate lumt given in Specification 3 3 A 2 h not casceJed. l'our radiation monitors are prosiJctl in stic reatior liuildmg semilation ducts whc.h imiiate isolation of the reactor builJmp and operation of the standby ras uc.itment system.1he monitors are im etcJ in the reactor buikimg sentil.itum dosa l he eisp loric as.: sme.out of two for e.nh sci, and eagle set can inicianc a trip anderemtent of the oeher sce Any up. sale inp u dt sause the demrd atrion leap sentmp of ? rnRrbt im amnon in the ventilation Jun s.nc hne d n em mon ihop not uul sentilinon nolition and standby rn uc.onn ne syutm ogwration e no elui tienenotation s M i(Ic.nc i.iic huus rn en in y sit' san.m ) E A 3 n not est scded.1mo eadiatum momtors \\ 9 are prinidcJ on the him hn;= stain whnh untuh nulanon of she scattor budstmr.mJ opci.ienm.if the standby gas ticatment systs na l he nip tores n ame.out of two 'liep wnmp of tilu mRrtu for the m.nnems on the refucione ~r.nc l swa op.m mio "mr a a m d ss md aian not oia md "andhy r-es acaom ai sF= "Peranaa n i 3.2/l.1 7 Amendment No. 61

i QUAD-CITIES ~ O so that none of the activity released during th'e refueling accident leaves the reactor building via the normal ventilation stack but that all the activity is processed by the standby gas treatment system. 9 The instrumentation which is provided to monitor the postaccident condition is listed in Table 3.2 4. The instrumentation listed and the limiting conditions for operation on these systems ensure adequate monitoring of the containment following a loss-of. coolant accident. Information from this instrumentation will provide the operator with a detailed knowledge of the conditions resulting from the accident; based on this information he can make logic.it decisions regarding postaccident recovery. The specifications allow for postaccident instrumentation to be out of service for a period of 7 days. This period is based on the fact that several diverse instruments a*e available for guiding the operator should an accident occur, on the low probability of an instrument being out of service and an accident occurring in the 7-day period, and en engineering judgment. The normal supply of air for the control room ventilation system comes from outside the service building. In the event of an accident, this source of air may be required to be shut down to prevent high doses of radiation in the control room. Rather than provide this isolation function on a radiation monitor installed in the intake air duct, signals which indicate an accident,'i.e., high drywell pressure, low water level, main steamline high flow, or high radiation in the reactor building ventilation duct, will cause isolation of the intake air to the control room. The above trip signals result in immediate isolation of the control room ventilation system and thus minimize any radiation dose. !O I i l 1 r 9 3.2 / 4.2-8 Amendment No. 61

QUAD-CITIES DPR-29 nb TABLE 3.24, gesTRUMENTAil0N TNAT Id!T!ATES ROD BLOCK Indman assest of operous er 1stesse suivament tasanets per Tate sritem'" lastrummest Me tafel htthg 2 APRM wscale (fbw tnasyn $[0.650W + 43') _ FRP D gpLpp 2 APRM upscale (Refuel and Statup/ Hot 512/125 fuD scale Standby mMe) 2 APRM downscale'n 2:3/125 fut scale 1 Rod bbck monttor escale (Bow biosyn so.650W + 425 1 Rod block monttor downscale" 23/125 fut scale t 3 glM downscale * * '23h25 fut scair 3 glM ispscale* $108/125 fv5 scak 2m SO4 detector not si Startup position

• 22 feet bebw core center.

Ire 3 ged detector not in Statup position

• 22 feet bebw core center-Ine 28
• 558 sqtscale 510' counts /sec 2"

' ~

O 1 Nigh water level si saam dscharge volume 525 galons 18otes

1. For the Startup/ Mot Standby and Hun positions of the reactor mode cc1cetcr switch, there shall be two ope able or tripped tr.ip systems for cac:t fune-tion except the SRK rod blocl;ts. IRM upscale and IRM downscale nued not be scale (flow binned),

operable in the kuri position, APitM downeesle, APRM up/ Hot Standuy mate. l and RnM down::cale need not be operabic in the Startup i The RTE 1 upscale riced riot be o[>erable ut less than 30% rated thermal power. One chanrici may be bypar:ned above 30% rated thermal power prov.3ded tha t a limitinf, control rod patter:: (toos not exist. For system: with more than one channel per trip system, if the first column cannot be met for one of the two trip systems, this condition may exist for up to 7 days provided $ hat ditrinr. th:st tituc the operabln sy tesi is functionally tr.:sted im-mediately and daily thereorter : if this condition lastu longer thuri 7 days the system chull be tripped. If the first column cannot be mot for both trip systems, the systems shall be tripped. -l

2. W is the percent of drive flow required to produce a rated core flow of g

9e raillion 1b/hr. Trip level setting is in percent of rated power (2511

• t).

1 su ammaste may to bypassed seen a e si es huest range. & he taunen a typensed seen its met rate e 2100 CPS. i ene af me les See seats not be typsened. 6 The Stilenctan met he bypasand a the Aq%er IRID fanget hanget 8.1 and 10) when ths flti spegels red tiett is egetable 7. Ibt eegeset to to spean ende poterest tea powe phpcs tests at atmoghers preenere dereg er ehm reteeleg at pe=w touts not to enceed 5 Mwt t The M luncten stros when the seater made sent:ns a the Relves er steetvoMet stendty peniten. t he e, e hvaseasd when ins sms a tea, esetad .: 3.2M2-14 Amendment No. 61

QOAD-CITil'.S I)l 'I'-29 . 2-v 1ASLE 3.24 POSTACClotNT MONITORING INSTRUMENTAkl0ll REQUIREMENTS 78 betrueest theisse go,ter Reedest of Operable Locaties Bester themsels'H 88 Peremeter Bell 1 Provided f ege i no.eiur pre wre 901-5 1 oi500 psia ? 01200ps4 i neector weier sevei 901-1 2 .ioo inches e 200 incem to inches is sop or tosis

• i Terve weier temper ~re 901-71 2

02o0* r ~ 1 Ton,e si, wmpo,eture 901-21 2 0600* - Terve eseter levoi. 901-3 1 25 inees - + 25 W I i"di"***' 2 *I Torve water level. 1 18 inch range eight glose i Torve presmere 901-3 1 4 inches He to 5 esis 1 Drywell preomste 901-3 1 4 incem Hg to 5 osse l 0 to 75 psig !-O or we t.em. ore ~,e 901-2i s o.eo - 8 2 Neutron nwnitoring 901-5 4 o.t.10 ces 84 3 l 2 Torus to drywell 2 03 psid differential pressure I i Itsees 1. Instrument channels required curing pos or operetton to rnordnor pasteccident conditions. t l 2. Proreione are made for accel sempling and monitoring of drywell etmosphere. I, i I, ,8

• Top of active fuel is defined to be 360 inches above vessel sero (See Bases 3.2).

i e, i O i E 3.2/42-15 y Amendment No'. 61

QUAD-CITIES IFR-29 O

3. The control rod drive housing support

3. The correctness of the control rod system shall be in place during reactor withdrawal sequence input to the power operation and when the reactor RWM computer shall be verined after coolant system is pressurized above loading the sequence.

atmospheric pressure with fuel in the reactor vessel, unless all contrni rods Prior to the start of control rod with-are fully inserted and Specircation drawal towards criticality, the capabil-3.3.A.I is met. ity of the rod worth minimizer to properly fulfill its function shall be Control rod withdrawal sequences verined by the fo!!owing checks: a. shall be established so that max-s. The RWM computer entine diag-imum reactivity that could be nostic test shall be successfully l ; added by dropout of any incre-ment of any one control blade performed. would be such that the rod drop accident

b. Proper annunciation of the seiec-

+ design limit of 280 cal /cm. is _not exceeded. tion error of one out-of-sequence l

b. Whenever the reactor is in the Startup/ Hot Standby or Run

c. The rod block function of the i

mode below 207. rated thermal RWM shall be verified by with-l power, tne rod worth minimizer drawing the 6rst rod as an out-l shall be operable. A second opera-of sequence control rod no more tor or qualified technical person than to the block point. may be used as a substitute for an inoperable rod worth minimizer which fails after withdrawal of at least 12 control rods to the fully l withdrawn position. The rod worth minimizer may also be l bypassed for low power physics testing to demonstrate the s!wl-down margin requiremero- ~ Specincation 3.3.A if a weln engineer is present s'# 9 n.;. step by-step rod mo m w. 9 sest procedure.

4. Control rods shall not he withdrawn 4.

Prior to control rod withdrawal for for startup or refueling unless at least startup or during refueling. verify that two source range channels have an at least two source range channels + observed count rate equal to or greater have an observed count rate of at least than three counts per second and these

).ree counts per second.

SRM's are fully inserted. lO S. During operation with limiting con-

5. When a limiting control rod' pattern trol rod patterns.as determined by the exists. an instrument functional test of nuclear engineer, either:

the RBM shall be performed prior to

a. both RBM channels shall be withdrawal of the designated rod (s) i operable.

and daily thereafter.

b. control rod withdrawal shall be blocked; or k

O 3.3/4.3.1 Amendment No. 61

QUAD-CITIES

• IVH-29 (G3 the operating power level shall be c.

( limited so that the MCPR will re - main above the MCFR fuel cladding integrity safety limit assuming a sin-gle error that results in complete withdrawal or ny single operable a control rod. C Scram insersion Times C Sersm insertion Times

1. The average wram insertion time, ha-I. Aller refueling outage and prior to sed on the deenergization of the scram operation above 30% power, with re-pilot valve solenoids at time sero.of all actor pressure above 800 psig. u!! con-operable control rods in the reactor trol rods shall be subject to scram time power operation condition shall be no measurements from the fully with-greater than drawn position.The scram times shall be measured without reliance on the Awrage Scram control rod drive pumps.

% inserted From insertion fully Withdrawn Times (sec) 5 0.375 20 0.900 50 2.00 90 3.50 O The average of the scram insertion times for the three fastest control rods of all groups of four control rods in a two by two array shall bc no greater than: % inserted From Average Scram Adfy Withdrawn Times (sec) 5 0.39M 20 0.954 50 2.12 90 3.80

2. The maximum scram insertion time

2. Fellowing a contro!!cd shutdown of for 90% insertion of any operable con-the reactor, but not more frequently trol rods shall not ruved 7 seconds.

than 16 weeks not less frequently than 32 wec) intervals 50% of the control

3. If Specification 3.3.C.I cannot he met.

9" the reactor shall not be made super-eritical: if operating. the reactor shall mam i mes spec ed y Spec.ncanon i be shut down immediately upon deter-3.3.C. Allc ntr I r d drives shall have mination thai uverage scram time is expertenced scram test measurements Arnim caeh year. Whenever all of the control 4. If Specification 3.3.C.2 tannot he met. rod drive scrum times have been mea-the deficient control rod shall be con-sured. an evaluation shall be made to O. 3.3/4.3-4 Amendment No. 61 n..

QUAD CITIES DPR-20 ~ B. Control Rod Withdrawal 1. Control rod dropout accidents as discussed in Reference 1 can lead l to s1 nificant core damacc. If coupling integrity is maintuined, 6 the possibility of a rod dropout accident is climinated. The over-travel position feature provides a positive check, as only uncoupled drives may reach this position. Nutron instrumeniation response to rod movement provides a verification th'st the rod is following its drive. Absence of such response to drive movement would indicate an uncoupled condition.

2. The control rod housing support restricts the outward movement of a control rod to leu than 3 snehes in the catremely remote event of a nousing failure The amount of reactivity which could be added by this small amount of rod withdrawa' which is test than a normal single witt.drawai increment, will nor contribute to any dernage to the primary coolant system lhe design basis is given in Section 6 6 I. and the design evaluation is given in Section 6 6.3 of the SAR. This support is not required if the resetor coolant system is at aimospheric pressure, since there would then be no drivint forer to rapidly eject a drive housing. Additionally, the suppon is not required if all control rods are fully inserted or if an adequate shutdown margin with one control rod withdrawn has been demonstrated, since the reactor would remain suberitical even in the event of complete ejection of the strongest control rod 3

Control rod withdrawal and insertion sequences are established to ascure that the maximum insequence individual control rod or control rod accments which are withdrawn could not be worth enough to cause the rod drop occident design limit of 280 cal /cm to be execeded if j they were to drop out of the core in the manner defined for the rod drop accident. These sequences are developed prior to initiril oper-ation of the unit following any refueling outare rtnd the req ui t t: ment (N that an operator fallow thene ncquences is stipervit:cct by t he HWM or ) a second qualified ctatlon employee. These acquencc:: are developed g s to limit reactivity worths of control rods and 3 'together with the integral rod velocity limiters and the action of the control rod drive s,vstem. limits potential reactivity insertion such that the results of a control rod drop accident will not euce d a minimum fuel energy content nr 250 cal /gm lhe peak fuel enilulpy of 280 cal /gm n beinw the energy content at which rapid fuel siispersal and primary sys em damage ha e been foucd to occur t>ased on esperimental dai. as is diwuned in Iteferenec 2. \\ Trie analpit of the control rod drop accident was originally presented in Sectient 7.9.3.14 21.1 and 1411.4 of the SAR Imprmements in analyucal capahahty hne allowed a rnore refined analysis of the control roJ drop acuJent. ,These techniques are described in a topical report (Reference 2) and two supplements (References 3 a nd 4 ). In addition, a banked position withdrawal sequence described in Reference 5 has been developed to further reduce incremental rod worths. Method niid hauja for the rod drop accident antalysco are documented in Reference 1. By using the analytiaal modcli dewribed in those reporn seupled with conservative or worst. case input paranwicts, et has been determine I that for power locli Icss th.in2tG of rated power, the l speci6ed hmit on insequence cunirol and nr control rod segment worths wil! fimit the peak fuel enthalpy to less than 280 cal /r Above20% power even smgle opes.unt errors cannot result in g out-of-sequence control rod moriht which are sumcient to reach a pent fuel enthalpy of 250 cal /t should a postulaicd control rod drop accident occur. The following parameters and worst-case assumptions have been utilir.ed in the analysis to determine compliance with the 280 cal /gm peak fuel enthalpy. Each core reload will be analyzed to show conformance to the limiting parameters,

s. an interassenibly local peaking factor (Referenet. 6).

l m 3 3/h.3-8 Amendment No. 61

QUAD CITIES nm-s

b. the delayed neutron fraction chosen for the bounding reactivity curve l

c. a ber' inning-of-life Doppler reactivity feedback

d. scrcim times slower than the Technical Spec'ification rod scram insertion I

l rate (Section 3 3.c.1)

e. the maximum possible rod drop velocity of 311 fps f the design accident 'nd scram reactivity shape funculon, and

g. the moderator temperature at which criticality occurs In most cases the worth of insequence rods or rod segments in conjunction l

with the actual values of the other important accident analysis parameters described above, would most likely result in a peak fuel enthalpy sub-l stantially less than 280 cal /g design limit. Should a control drop accident result in a peak fuel energy content of 280 cal /g. fewer than 660

7) f O rods are conservatively estimated to perforate.This would result in an offsite dose well below the guideline value of 10 CFR 100. For 8 x 8 fuel, fewer than 850 rods are conservstively estim to perforate, with nearly the same consequences as for the 7 x 7 fuel case because of the rod pow differences.

The rod worth minimizer provides automatic supervision to assure that out of sequence control rods will not be withdrawn or inserted; i.e it limits operator deviations from planned withdrawal sequences (reference SAR Section 7.9). It serves as a backup to procedural control of control rod worth. In the event that the rod worth minimizer is out of service when required. a licensed operator or other qualified technical employee can manually fulnll the control rod pattern conformance 9 function of the rod worth minimizer. In this case, the normal procedural controls are backed up by independent procedural controh to assure conformance.

4. The source range monitor (SRM) system performs no automatic safety system function,i.e., it has no scram function. It does provide the operator with a visual indication of neutron level. This is needed for knowledgeable and efficient reactor startup at low neutron levels. The consequences of reactivity accidents are functions of the initial neutron flux.The requirement of at least 3 counts per second assures that any transient, should it occur, begins at or above the initial value of 10 ' of rated power used in the analyses of transients from cold conditions. One operable SRM channel wou adequate to monitor the approach to criticality using homogeneous patterns of scattered control ro withdrawal. A minimum of two operable SRM's is provided as an add:d conservatism.

5. The rod block monitor (RBM) is designed to automatically prevent Ibei damage in the event of erroneous rod withdrawal from locations of high power density during high power operation. Two channels are provided, and one of these may be bypassed from the console for maintenance end/or '

testin5. Tripping of one of the channeb will block erroneous rod withdrawal soon enough to prevent fuel damage. This system backs up the operator, who withdraws control rods according to a written sequence. The specified restrictions with one channel out of service conservatively assure tn damage will not occur due to rod withdrawal errors when this condition exists. During reactor the withdrawal of operation with certain limiting control rod patterns, more fuel rods wi th i a designated single control rod could result in one or, ladding integrity safety limit.During use o MCPh's less than the MCPR fuel it is judged that testin,of the Rb 4 system to assure its operability prior to withdrawal will assure tha' :.aproper withdra wat does not occur. It is the responsibility of the Nuclear En to identify these limiting patterns and the designated rods either when the patterns are h estabbshed or as they develop due to the occurrence ofinoperable control rods in othr, than patterns. 3 3/4.3-9 Amendment No. 61

QUAD CITIES OPR-29 l C. Scram Insertion Times The control rod system is analyzed to bring the reactor subcritical at a rate fast enough to prevent fuel damage, i.e., to prevent the MOPF. from becoming less than the fuel cladding integrity safety limit. i j A'nalysis oT the' limit 3.ng power transient shchs that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the above specification, provide tr.e reoutred protection, and MCPR remains 6reater than the fuel cladding inte5rity I safety limit. The mimimum amount of reactivity to be inserted durino a scram is controlled by permitting no nnre than 107, of tha cperable rods to have lorri scram times. In the analytical treatment of the transients, 290 milliseconds are l allowed between a neutron sensor reaching the scram noint and the start of motion of the control rods. This is adequate and conservative when corrpared to the typically observed time delay of about 210 milliseconds. Approximately 90 l milliseconds after neutron flux reaches the trip point, the pilot scram valve solenoid deenergizes and 120 milliseconds later the control rod motion is estimated to actually begin. However, 200 milliseconds rather than 120 milliseconds is conservatively assumed for this time interval in the transient analyses and is also included in the allowable scram insertion times specified in Specification 3.3.C. The scram times for all control rods will be determined at the time of each refueling outage. A representative sample of control rods will be tested following a shutdown. Scram times of new drives are approximately 2.5 to 3 seconds; lower rates of change in scram times following initial plant operation at power are expected. The test schedule provides reasonabic assurance of detection of slow drives before system detarioration beyond limits of Sriecification 3.3.C. Tho program was developed on tho banis of the statistical approach outlinal below and judqment. 1 I. i Tbc history of drive performance accumulated to date indicates that the 90% insertion times of new and overhauled drives approximate a normal distribution about the mean which tends to become skewed toward lonyt scram times as operating time is accumulated.The probability of a drive not exceeding the mean 90% insertion time by 0.75 seconds is greater than 0.999 for a normal distribution. The l measurement of the scram performance of the drives surrounding a drive exceeding the expected range of scram performance will detect local variatior.s and also provide assurance that local scram time hmits are not execcded. Continued monitoring of other drives exceeding the expected range of scram times provides surveillance of possible anomalous performance. The numerical values assigned to the predicted scram performance are based on the analysis of the Dresden 2 startup dria and of data from other llWlUs such as Nine Mile Point and Oyster Creek. l ') The occurrence of scram times within the limits. but signifuntly longer than average, shnuld b: viewed as an indication of a systematic problem with conttol rod drives, especially if the number of drives j cahibiting such scram times cacceds eight. the allowable number ofinoperabic rods. l 3.3 / 4.3-10 l Amendm5nt No. 61

QUAD-CITIES ,o DPR-29 The occurrence of scrarri times within the limits. but sienificantly longer than average. should be view ~l as an indication of a systematic problem with control rod drives, especially if the number of drives exhibiting such scram times exceeds eight. the allowable number ofinoperable rods. D. Control Rod Accumlators The basis for this specification was not described in the SAR and is therefore presented in its entire Requiring no more than one inoperable accumulator in any nine-rod square array is based on a series of XY PDQ 4 quarter core calculations of a cold clean core. The worst case in a nine-rod withdrawal sequence resulted in a L,, < l.0. Other repeating red sequences with more rods withdrawn resulted in k,, > 1.0 At reactor pressares in excess of 800 psig, even those control rods with inonerable accumulators will be able to meet required scram insertion times due to the action of reactor pressure. In addition, they may be normally inserted using the control rod drive hydraulic system. Procedural control will assure that control rods with inoperable accumulators will be spaced in a one in nine array rather than grouped together. E. Resethity Anomalies During each fuel cycle. excess operating reactivity varies as fuel depletes and as any burnable poison in supplementary controlis burned. The magnitude of this excess reactivity may be inferred from the critical rod configuration. As fuel burnup progresses anomalous behavior in the excess reactivity may be detected by comparison of the critical rod pattern selected base states to the predicted rod inventory at that state. Power operating base conditions provide the most sensitive and direcily interpretable data relative to core reactivity. Furthermore, using power operating base conditions permits frequent reactivity compartsons. Requiring a reactivity comparison at the specified frequency assures that a comparison will be made before the core reactivity change exceeds 1% ak. Deviations in core reactivity greater than 1%.1k are 1 not expected and require thorough evaluation. A 1% reactivity limit is considered safe since an insertion of the reactivity into the core would not lead to transients exceeding design conditions of the r; actor system. I F. Economic Generation Control System Operation of the f.icility with the economic generation control system (EGC)(automatic flow control) is limited to the range of 65% to 100% ofrated core flow. In this flow range and with reactor power above 20% the reactor could safely tolerate a rate of change ofload of 8 MWe/sec (reference SAR Section 7.3.5 ). Limits within the EGC and the fiow control system prevent rates of change greater than appro 4 MWe/sec. When EGC is in operation, this fact uill be indicated on the main control room console. The results ofinitial testing will be provided to the NRC before the onset of routine operation with EGC. References

1. " Generic Rel,ad Fuel Applica tion", NEDE-214011-P-A"

2. C. J. Paone. R. C.Stirn, and J. A. Wooley.* Rod Drop Accident Anal) sit for Large BWR *v, GE Top NEDO.10.527. Marsh 1972

3. C. J. Panne. It. C.Siirn, and R. M Young.* Rod Drop Acciden: Analysis for Large BWR*f Supplement GE Topi:al Report Nr.DO.10527 July 1972.

4. J. M. Haun. C. J. Paone, and R. C. Stirn. ' Rod Drop Acciden Analysis for Large BWR's, Addend Emposed Cores,' Supplement 2. GE Topical Report NEDO.10327, January 1973.

5. C. J. Paonc, " Banked position withdrawal sequence," Licensing topical Report 'NEDO-21231, Janua ry, 1977.

6. To include t.ne power spike errect cou::ed by gaps between ruel pellets.

• Approv'ed revini on rmmbe r a t, t.i me reload fuel orial.vr.sta are l r r..nne.l.

j 3.3/4.1-11 l t l. Amer]dment No. 61 t L ~r

QUAD CITIES DlH-29 i 3.4 LIMITING CONDITIONS FOR OPERATION DASES p%) A. The design objective of the standby liquid control system is to provide the capability of bringing the reactor from full power to a cold, xenon-free shutdowrt assuming that none of the withorawn control rods can be inserted. To meet this objective, the liquid control system is designed to inject a quantity of boron which produces a concentration of no less than 600 ppm of boron in the l reactor core in approximately 90 to 120 minutes with imperfect mixing. A borort concentration of 600 ppm in the reactor core is required to bring the reactor from full power to 3% Ak or more ) subcritical condition considering the hot to cold reactivity swing, menon poisoning and an additional margin in the reactor core for imperfect mixing of the g chemical solution in the reactor water. A normal quantity of 3470 gallons of solution having a 13.4% sodium pensaborate concentration is required to meet this shutdown requirement. The time requirement (90 to 120 minutes) for insertion of the boron solution was selected to override the rate of reactivity insertion due to cooldown of the reactor following the xenon ponon peal For a required pumping rate of 39 gpm. the maximum storage volume of the boron solution is estabhshed as 4875 gallons (195 gallons are contained below the pump suction and, therefore. cannot be inserted). Boron concentration, solution temperature. and volume are checked on a frequency to assure a fin reliability of operation of the system should it ever be required. Experience with pump operabihty indicates that monthly testing is adequate to detect if failures have occurred, ne only practical time to test the standby liquid control system is during a refueling outage and by initiation from local stations. Components of the system are checked periodically as described above and make a functional test of the entire system on a frequency of less than once each refueling outage g unnecessary. A test of explosive charges from one manufacturing batch is made to assure that the charges are satisfactory. A continual check of the firing circuit continuity is provided by pilot lights in the control room. B. Only one of the two standby liquid control pumping circuits is needed for proper operation of the system. If one pumping circuit is found to be inoperable. there is no immediate 6 teat to shutdown capability, and reactor operation may continue while repairs are being made Assurance that the remaining system will perform its intended function and that the teliability of the system is good is obtained by demonstratsng operation of the pump in the operable circuit at least once daily. A reliability analysn indicates that the plant can be operated safely in this manner for 7 days. C. The solution saturation temperature of 13% sodium pentaborate, by weight,is 59' F.The solution shall be kept at least 10' F above the saturation temperature to guard against boron precipitation. The 10' F margin is included in Figure 3.3 1. Temperature and liquid level alarms for the system are annunciated in the control room. Pump operability is checked on a frequency to assure a high reliability of operation of the system should it ever be required. Once the solution has been made up. boron concentration will not vary unicss more bornn or more water is added. Levelindication and alarm indicate whether the solution volume has changed, which might indicate a posuble solution concentration change. Considering these factors, the test interval has been established. 3.4/4.4-3 t Amendment No. 61

QUAD-CITIES DPR-29 is being done which has the potential for draining the reactor vessel.

3. When irradiated fuel is in the reactor and the vessel head is removed, the suppression chamber may be drained completely and no more than one con-trol rod drive housing opened at any one time provided that the spent fuel pool gate is open and the fuel pool water level is maintained at a level of greater than 33 feet above the bottom of the pool. Additionally, a minimum condensate storage reserve of 230,000 gallons shall be maintained, no work shall be performed in the reactor vessel while a control red drive housing is l

blanked following removal of the con-trol rod drive, and a special nange shall be available which can be used to blank an open housing in the event of a leak. 4. When irradiated fuel is in the reactor and.the vessel head is removed, work that has the potential for draining the j vessel may be carried on with less than 112.200 ftSpfwater in the suppression pool, provided that: (1) the total vol-ume of water in the suppression pool, l refueling cavity, and the fuel storage pool above the bottom of the fuel pool gate is greater than 112,200 IV; (2) the fuel storage pool gate is re-i moved;(3) thelow-pressurecoreand containment cooling systems are oper-able; and (4) tne automatic mode of l, the drywell sump pumps is disabled. G. Maintenance of Filled Discharge Pipe G. Maintenance of Filled Discharge Pipe The following surveillance requirements shall be adhered to to a'ssure that the discharge piping of the core spray, LPCI mocle of the RHR, HPCI, and RCIC are Alled:

1. Whenever core spray, LPCI mode of

1. Every month prior to the testing of the the RHR, HPCI, or RCIC are required LPCI mode of the RHR and core spray to be operable, the discharge piping l

ECCS, the discharge piping of these from the pump discharge of these sys-systems shall be vented from the high tems to the last check valves shall be point and water now observed. Alled. Amendment No. 61 3.5/ 4.5-7

QUAI)-CITIES DPR-29 -) cycle by assuring that water can be run th ough the drain knes and actuating the ait-operated vahes by operation of the following sensors:

1) loss of air

2) crjuipment drain sump hirh level

3) vault high Icvel

d. The rondenser pit 5-font trip cii-cuits for cadi ch.innel shall be checLed once a month. A logic system functional test shall be per-formed during each refueling outage.

I. Average Planar LliGR L Average Planar LHGR During steady. state power operation, the average lir.sar heat generation rate (APLHCR) of all the Daily during steady state operatiJn rods in any fuel assembly,as a function of avenage above 2'3% rated thermal power' planar exposure, at any axial location, shall not the ayerage planar LHGR shall be determined. exceed the maximum average planar LilGR shown in Figure 3.51 If at any time [ 9 during operation it is determined by nurinal sur. weillance that the limiting value for APLilGR is J. Leest LHC1 being exceeded, action shall be initiated within 15 minutes to restore operatsun to.rithin the pre. Daily during steady-state power operation scribed lirnats. If the APLIIGR is not returned in above 25'6 of rated thermal power, the local within the prescribed limits within 2 hours, the LHGR shall be determined. I reactor shall be brought to the cold shutdown condition within 16 hours. Survedlance and correspondmg action shall continue unta reactor operation is within the presenhed limits. J. Imcal LHGR During steady. state power operation, the linear heat generation rate (LilGR) of any rod in any fuel assembly at any axial location shall not exceed the maximum allowable LilGR. If at any tirne during operation it is determined by normal survedlance that the hnuting value for LilGR is being exceeded, action shall be initiated within l$ minutes to restore operation to within the pre-scribed limits. If the LilGR is not returned to O Amendment No. 61 3'5"$# H

QUAD CITIES nPn-29 within the prescribed limits within 2 hours, the { O s i> 8 6:eue t 'o ta 'o su'd -n s condition withm 36 hours. Surveillance and coe. re8Ponding action shall contmue untd reactor

• Peration is withm the prescribed limits.

Maximum allowable LHGR for all 8x0 fuel types is 13.4 KW/ft. For 7X7 and mixed oxide fuel, the neximum allowable LilGR is as follows: c n LHG R_, < LilG R, 1 -( J P/ P ),,,,( L/ L, ) -J where:

LHGR,

=, design LHGR 17.5 kW/fi. = ( AP/PL, = manimum power spiking penulty .035 initial core fuel ') = .029 reload I, 7 x 7 fuel = i =.028 reload 1,7 x 7 mixed oxide fuel L, = total core length = 12 feet Asial distance from hottom ol' core L = K. Minimum Critical Power Ratio (MCFR) K. Minimum Critical Puner Ratin (MCPR) During steady. state operation MCPR shall he The MCPR shall he determined daily Jurint greater than or equal to steady-state power operation.ihme 25% of 1.35 (7 x 7 fuel) rated thermal pow. r. 1.35 (8 x 8 fuel) st rated power and flow. If at any time during operation it is determined by normal survedlance that the limiting value for MCPR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. if the steady. state MCPR is not returned to within the prescribed limits within 2 hours, the reactor shall be brought to the cold shutdown condition within 36 hours. Suiveillance and corresponding action shall_ contmue until reactor operation is within the prescribed hmits. For core flows other than rated, these nominal values of MCPR shall be increawd by a facto of kg where k i g s as shown in Figure 3.5.2. O. 3.5/ 4.5-Io Amendment No. 61

QUAD-CITIES DPR-29 'l ..5 LIMITING CONDITION FOR OPERATION DASES A. Core Spray and LPCI Mode of the RIIR Spiem This specification assures that adequate emergency cooling capability is available whenever irradiated fuelis in the reactor vessel. . Based on the loss.of coolant analytical methods described in General Electric Topical Report NEDO.20566 and the specific analysis in Reference 1, core cooling systems provide sufficient cooling to ~he core to dissipate the energy associated with theloss-of. coolant accident,tolimit calculated perature to less than 2200*F, to assure that core geemetry remains intact, to limit cladding meta t action to less than 1%,and to limit the calculated !ocal metal-water reaction to less than 171 h The limiting conditions of operation in Specifications 3.5.A.I through 3.5.A.6 specify the combi of operable subsystems to assure the availability of the minimum cooling systems noted a failure of ECCS equipment occurring during a loss of-coolant accident under these limiting condit of operation will result in inadequate cooling of the reactor core. Core spray distribution has been shown, in full-scale tests of systems similar in design to Quad. Cities I and 2, to exceed the minimum requirements by at least 25% In addition, coolin ectiveness has been demonstrated at less than half the rated flow in simulated f _ beater rods to duplicate the decay heat characteristics,of irradia efr puuure has fallen to 90 psig. The LPCI mode of the RilR system is designed to provide emergency cooling to the core by floodin the event of a loss-of. coolant accident.This system functions in combination with the core spray sysicm l to prevent excessive fuel cladding temperature. The LPCI mode of the RHR system in co d the core spray subsystem provides adequate cooling for break areas of approximately 0.2 ft' up to the latter being the double. ended recirculation line break with the equalizer line including 4.18 ft), ~ li g 'uerseen th[recircti!ation loops closed without assistance from'the high. pressure emergency core coo n i subsystems. The allowable repair times are established so that the average risk rate for repair would be n the basic risk rare.The method and concept are described in Reference 3 Using the results develope l this reference, the repair period is found to be less than half the test interval. This assumes that the core spray subsystems and LPCI constitute a one.out.of-two system; however, the combined e systems to limit excessive cladding temperature must also be considered. The test interv Specification 4.5 was 3 months. Therefore, an allowable repair period which maintains th considerinjr single f.ulures should be less ih.m 30 days, and this spectfication is within this perio multiple failures, a shorter intervalis specihed; to improve the assusanec that the semainin ihnetion, a daily test is called for. Although it is recognized that the information given in Reference provides a quantitative method to estimate allowable repair tim stated in the specific items were established with due regard to judgment. Shouid one core spray subsystem become inoperable, the remaining core spray subsystem a I.PCI mode of the RHR system are available should the need f6r core cooFng arise: To assure th . remaining core spray, the LPCI mode of the RilR system, and the diesel generators are available are demonstrated to be operable immediately. This demonstration includes a manual initiation of the pumps and associated valves and diesel; enerators. Based on judgments of the relia systems,ie., the core spray and LPCI, a 7 day repair period was obtained. I3 i q I I i t 3.5/4.5 11 bl. I Anendment No. 61 >,.), J -. ~.

CUAD CITTIS DPR-29 r H. Condensate Pump Room Flood Protection See Specific.ition 3.5.H. I. Aserage Planar f.HGR This specification assures that the peak cladding temperature following the postulated design-basis lossof coolant accident wd! not exceed the 2200 F limit specified in the 10 CFR 50 Appendut K considenng the postulated effects of fuel pellet densification. The peak cladding temperature following a postulated lossof<oolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any anallocation and is only secondarily dependent on the rod.to rod power distribution witidn an assembly. Since expected local variations in power distribution withir' a fuel assembly affect the calculated peak cladding temperature by less than 120*F relative to the peak temperature for a typical fuel design, the hmit on the average planar U!GRis sur-ficient to assure that calculated ternperatures are below the hmit. The maximum average planar LHGR's shown in Figure 3.51 are based on calculations employing 'he models described in Reference 2. J. localI.HGR This specification assures that the maximum linear heat generation rate in any rod is few than the d linear heat generation rate even if fuel pellet densification is postulated. The power spike penalty As discussed in Reference 2 and assumes a linearly increasing variation in axial l gaps between core bottom and top and assures with a 95Fo confidence that no more than one fuel exceeds the design linear heat-generaticn rate due to power spiking. l K. Minimum Critical Power Ratio (MCPR) The steady state values for MCPR specified in this specification were selected to provide margin to accommo-dat'e tranuents ai.d uncertainties in monitoring the core operstmg state as well as uncertainties in the entical power correlation itself. These values also assure that operation will be such that the initial condition assumed j For j for the IDCA analysis plus two percent for uncertainty is satisfied. sny of the special a:t of uansients or disturbances caused by single og<rster error or snele equipment malfunction,it is required that des:en analyses mituhted at this steady. state operating hmit yield a MCPR of not his than that necified in SpeciScatien 1.1.A at nr.) tline during the tranuent assuming instrument tnp sett:ms gMn m Specifiestion 2.1. For sn:Jys:s of the f the value of MCPR stated in this thermal consequences of th2se transients, specification for the limiting condition of operation bounds the initial prior to the initiation of the transien'.s. value of MOPR assu ned to exist This initial condition, which is used in the transient analyses, will cre-clude violation of the fuel cladding inte6rity safety limit. Assumptiona fer cnd methods used in calculating the required steady state MCFR li tit The results apply with cach reload cycle are documented in Reference 2. increased conservatism while operating with MCFR's greater than specified. The most hmitmg teansients with respect to MCPR are geocraDy: a) Rod withdrawal error b) Imad sejection c Turbine Tsip without bypass c) 1masof feedwates heate: ! a r. es t factors influence which of these transients results in the Severa' reduction in critical power ratio such as the specific fuel loading, ex-1 The current cycles reload licensing analyst c spec-posure, and fuel type.iries the limitine transients for a given exposure increment for each fuel Tr,c values specified as the Limiting, Condition of Operation are con-restrictive over the entire cycle for i type. servatively chosen to t)ound the most i each fiael type. i 1 + ~l. V4. ">- 14 Amendment No. 61 i,

r QUAD-CITIES a .DPR. 29 For cose fbw tates less than rated, the steady state MCPR h inerened by the formula given in the i,,ceire sation. h auuies that the MCPP will be mamistned treates than that specified in Speification i I. A enn in the evenit tiist tiet niotut generator set speed controller etuses the scoop tube poutioner fut the ibJ cv.:pler to move to the maaamum speed positson. References

1. " Loss-of-Coolant Analysis' Report for Dresden Units 2, 3 and Quad Cities Units 1, 2 Iluclear Power Stations," HEDO-24146Aa, April, 1979

2. " Generic Reload Fuel Application," NEDE-24011-P-A**

3. I. M. Jacobs and P. W. Marriott, GE Topical Report APED 5736,

" Guidelines for Determinine Safe Test Int.ervals and Hepair T;.as for Er.61neered Safeguardc," April,1969

• Approved revision at time of plant operation.
  • Approved revision number at time reload fuel analyses are performed.

1 l i i k i o .. t t t 3.5/4.5-15 ) t Amendment No. 61

(/ QUAI)-CITIF.S DPR-29 i Should the switches at levels (a) and t h) f.sil or the operator fail to trip the circulating water pumps on alarm at level (b), the actuation of either level switch pair at level (c) shall trip the ciiculatine water pumps automane.ifly and alarm in the control room These redundam lesci switth pairs at levelle) are designed and installed to IH.I' 279. *Critetia for Nucle.ar Power I'lant Proiection Sperms' As the cirtulating water pumps are inpped. enhei manually or automatically. at level te) of 5 feet, the maumum water level re.iched in the condensrr pit due to pumping will be at elevanon $68 feet 6 inches elevation (10 rect above condenser pit floor elevation 558 feet 6 inches; 5 feet plus an additional 5 feet attributed to pump coastdown). In order to prevent the RIIR service water pump motors.md diesel. generator tooling water pump motors from overheatin

• a vault cooler it supphed for each pump l.ath vault cooler is desirned to inaintain the vault at a maximum 105* F temperature durmr operation or us rc pective pump. I or example. if diesel genes tror toolmr 1/2 3903 starts, in tooler also uarts and mamtains the vault at 105' F by removing heat supphe i water pump to the vault by the motor orpump i/2 3903. If.at the same time that pump I/2 3903 is in operanon. RilR sen ne water pump IC starn. its tooler will also start and compensate for the added heat supplied to the vault by the IC pump motor keeping the vault at 105' F.

Each of the coolers is supplied with cooling water from in respective pump's discharge line. After the water has been passed through the cooler it returns to its respectise pump's suction line. In this way the vault coolers are supplied with cooling water tusally inside the v.ault The cooling water quantity needed for each cooler o approximately I"r to 5% of the deugn flow of the pumps so that the recirculanon of this small amount of heated water will not affect pump or cooler operation. Operation of the fans and coolers is required during shutdown and thus additional surveillance is not required. Watertight vauhs for the ECCS pumps in the reactor building are tested in es>cntially the same manner and frequency as described for the condenser pump room vauln. Verification that acceu doors to cash vault are closed following entrance by personnel is covered by station operating pruteduret The LHGR shall be checked daily to determine if fuel burnup or control rod movement has caused changes in power distribution. Since changes due to burnup are slow and only a few control rods are moved daily, a daily check of power distribution is adequate. Aserage Planar UlGR At core thermal power levels less than or equal to 25%. operating plant experience and thermal hydraulic analpes indicate that the resuhm; aver. ige planar LilGR is below the maumum average planar LilG R by a considerabic margin; therefore, evaulaimn of the average planai I.llGR below this power lesel n not necessary. The daily requirement for calculanng avesare planar LilGR above 25% rated thermal pewc: is surheient, unce power distribution shifts are slow when there have not been sigmticant power or control rod changes. lacal LIIGR The LHGR as a function of core height sh'all he checked d.iily during tractor operation at greater than or equ.it to 25% power to determine if fuel burnup or conitol rod movement has caused changes in power distribution. A limiting LHGR value is precluded by a considerable margin when employing any permissible control rod pattern below 25% rated thermal power. Minimum Critical Pnact Hatio (MCPR) At core thermal power levels less than or equal to 25%. the reactor will be operating at minimum recirculation pump speed and the moderator void content will he very small. For all designated control rod patterns which may be employed at this point operating plant experience and therm.il hydraulic analyus indicate that the resulung MCPR value is in exceu of requiremenn by a conuderaNe margm. With this low void content, any inadvertent i j core flow inescate would only pl.nc opeiation in a mure comervative mode relative to MCPR. I l G l 3.5/4.5-18 Amendment No. 61 A

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.a; 5 0 100'00 200'00 3o000 ,9 PLANAR AVERAGE EXPOSURE (MWD /P) U \\ x l H Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) vs. Planar Ave. rage Exposure l i 4 ( heet 6 of 6) Amehdment No. 61 l.- L-1 )

r_ 4 s p aeog o,, UNITED STATES e 8 NUCLEAR REGULATORY COMMISSION o h WASHINGTON, D. C. 20555 ~g s,, s

• SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 61 TO FACILITY OPERATING LICENSE N0. DPR-29 COMMONWEALTH EDIS0N COMPANY AND IOWA-ILLIN0IS GAS AND ELECTRIC COMPANY QUAD CITIES NUCLEAR POWER STATION, UNIT NO. 1 DOCKET NO. 50-254 Introduction By letter dated September 2,1980 (Ref.1), and supplemented by letter dated October 3,1980 (Ref. 2), Comonwealth Edina Company (CECO or the licensee), proposed an amendment to Quad Cities Unit 1 Appendix A, Technical Specifications. CECO has proposed the: amendment to support its review of future reloads-for Quad Cities Unit 1 under the provisions of 10 CFR 50.59.

Our approval is only for the proposed amendment and does not constitute approval of future reloads under the provisions of 10 CFR 50.59. Evaluation Safety Limit Minimum Crfttical Power Ratio (SLMCPR) This change prcvides SLMCPR values in the Technical Specifications for all currently approved core loadings. With retrofit 8x8 fuel in the core the SLMCPR limit is specified as 1.07. Without retrofit 8x8 fuel, the SLMCPR is 1.06. These limits have previously been found to be acceptable for this use in Reference 3 and on this basis the proposed change is acceptable. Rod Drop Accident (RDA) Design Limit The RDA design limit has been modified from 1.3%a maximum rod worth to 280 cal /gm peak fuel enthalpy rise. The 280 cal /gm design limit is acceptable per Standard Review Plan NUREG 75-087. Also, the power level below which the rod worth minimizer is required was increased from 10% to 20% of rated power. This is conservative by comparison to the previous specification, is consistent with reactor safety analyses, and is acceptable. 8012300 MS

. Maximum Average. Planar Linear Heat Generation Rate (MAPLHGR) New MAPLHGR curves reflecting the improved flooding characteristics of retrofit 8x8 fuel have been proposed by the licensee. Curves for 8x8, 8x8 retrofit, and 7x7 fuel of the various enrichments anticipated for future Quad Cities 1 reloads and extending to burnups of 40,000 mwd /t have been proposed (References 1 and 4). The new curves are based on an assumed fuel loading with 156 retrofit assemblies. Any reload with fewer such assemblies will be nonconservative with respect to the analyzed case and therefore outside the scope of this approval. Based on our previous approval of MAPLHGR curves reflecting 8x8 retrofit fuel reflood characteristics (Reference 5) and extension of burnup to 40,000 mwd /t (Reference 6), the licensee's proposed changes are acceptable. Power Peaking The licensee has proposed to adjust the Average Power Range Monitor (APRM) amplifier gain based on the Maximum Fraction of Limiting Power Density (MFLPD). Such an adjustment would be made in the event of operation with a MFLPD greater than the Fraction of Rated Power (FRP), with the objective of preventing the fuel cladding integrity safety limits from being exceeded during anticipated operational transients. This adjustment will be applied above 25% rated thermal power which is consistent with the LHGR surveillance requirements and the Standard. Technical Specifications. Previously this objective has been met by reducing the APRM trip settings through multiplication by the ratio of the Limiting Total Peaking Factor (LTPF) to the Total Peaking Factor (TPF). Such a reduction in set points is required in the event of operation with TPF>LTPF. We have concluded that the maximum reactor power which could be attained during anticipated operational transients with the proposed APRM gain adjustment would be no greater than would be attained with the current procedure for adjusting APRM setpoints. This conclusion is based on the equivalence of the ratio FRP/MFLPD to the ratio LTPF/TPF, and can be explained as follows. The LTPF can be expressed as the design linear heat generation rate divided by the plant rated thennal power per unit length of fuel rod. In a similar manner the TPF can be expressed as the maximum linear heat generation rate divided by the plant operating power per unit length of fuel rod. From these definitions it is easily determined that the ratio LTPF/TPF is the ratio of the design linear heat generation rate to the maximum linear heat generation rate times the fraction of rated thermal power, or 1/MFLPD*FRP. Thus FRP/MFLPD and LTPF/TPF are equivalent.

. However, instead of multiplying the APRM set points by FRP/MFLPD the same result can be achieved by multiplying the APRM reading by MFLPD/FRP to get If the reactor is operating in a steady a gain-adjusted APRM reading (.before gain adjustment) is equal to FRP. state mode the APRM reading Therefore by adjusting the gain until the APRM reading is equal to MFLPD, the APRM reading has effectively been multiplied by MFLPD/FRP as required. To summarize, the proposed formulation does not involve a reduction in margin to the trip point, and eliminates the need for different limits for different fuel types. In addition adjusting the APRM gain is much easier than changing the APRM trip setting, so that there is less chance for human error. Reactor Protection System (RPS) Delay Time The licensee has proposed to change the RPS delay time from 100 to 50 msec (time from opening of the sensor contact up to and including the opening of the trip actuator contacts). This change stems from an inconsistency which has existed between the Technical Specification value of 100 msec and the 50 msec value assumed by General Electric in the licensing analysis. The licensee has confirmed that the procedures used for determining RPS delay time are consistent with the General Electric use and definition of a 50 msec delay time in the licensing analysis. The staff has confirmed that the licensee has in place the capability for demonstrating compliance with the more restrictive specification. The proposed change is acceptable. Typographical Corrections and Clarification of Bases The remaining changes fall into the category of typographical corrections and clarification of bases and do not, as such, represent a significant safety concern. Environmental Consideration We have detennined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this detennination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and pursuant to 10 CFR Section Sl.5(d)(4) that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment. Conclusion We have concluded, based on the considerations discussed above, that: (1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment

does n'ot involve a significant hazards consideration, (2) there is reasonable by operati,on in the proposed manner, and (3)public will not be endangered assurance that the health and safety of the such activities will be con-ducted in compliance with the Commission's regulations and the issuance of the amendment will not be inimical to the coninon defense and security or to the health and safety of the public. l Dated: December 5,1980 t r 1 E i r I t i I b ~

=.-. References l 1. Letter from R. F. Janecek (Ceco) to Director of Nuclear Reactor Regulation (.USNRC), dated September 2,1980. 2. Letter from R. F. Janecek +(CECO) to Director of Division of Licensing (USNRC), dated October 3,1980. 3. Letter from D. G. Eisenhut (USNRC) to R. Gridley (GE) dated May 12, 1978. 4. " Loss-of-Coolant Accident Analysis Report for Dresden Units 2, 3, and Quad Cities Units 1, 2 Nuclear Power Station'," NED0-24146A, dated April 1979. 5. Letter from D. L. Ziemann (NRC) to Cordell Reed (CECO), dated April 24, 1979. 6. Letter from T.' A. Ippolito (NRC) to D. L. Peoples (CECO), dated December 28, 1979. l L i l l r i f Mv'Wv-w,,,, " i,,

.rI 7590-01 UNITED STATES NUCLEAR REGULATORY COMMISSION DOCKET NO. 50-254 COMMONWEALTH EDISON COMPANY AND l IOWA-ILLIN0IS GAS AND ELECTRIC COMPANY NOTICE OF ISSUANCE OF AMENDMENT TO FACILITY OPERATING LICENSE f The U. S. Nuclear Regulatory Commission (the Commission) has issued Amendment No. 61 to Facility Operating License DPR-29 issued to Commonwealth Edison Company and Iowa-Illinois Gas and Electric Company, which revised the Technical Specifications for operation of the Quad Cities Nuclear Power Station, Unit No.1, located in Rock Island County, Illinois. The amendment becomes effective as of the date of issuance. This amendment (1) authorizes changes to the plant Technical ] Specifications by revising the Minimum Critical Power Ratio Safety Lim'it to apply to new fuel types, (2) modifies the Rod Drop Accident Design Limit from 1.3%a maximum rod worth to 280 calories / gram peak fuel enthalpy rise, (3) approves the use of new Maximum Average Planar Linear Heat Generation Rate curves reflecting 8x8 retrofit fuel reload charac-teris' tics and extension of burnup to 40,000 megawatt days per short ton, (4) replaces the Limiting Total Peaking Factor with the Maximum Fraction of Limiting Power Density for adjustnent of the APRM flux scram and rod block trip settings, and (5) cnanges the Reactor Protection System Delay Time from 100 to 50 milliseconds for consistency with the licensing analysis. All other changes correct typographical errors and clarify the basis. 8012300ff5

'b f, 7590-01 The application for the amendment complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appro-priate findings as required by the Act and the' Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amend-ment. Prior public notice of this amendment was not required since the amendment does not involve a significant hazards consideration. The Commission has determined that the issuance of this amendment will not result in any significant environmental impact and that pursuant to 10 CFR Section 51.5(d)(4) an environmental impact statement, ) negative declaration, and environmental impact appraisal need not be prepared in connection with issuance of the amendment. i For further details with respect to this action, see (1) the appli-cation for amendment dated September 2,1980, as supplemented October 3, 1980, (2) Amendment No. 61 to License No. DPR-29, and (3) the Commission's related Safety Evaluation. All of these items are available for public inspection at the Comission's Public Document Room,1717 H Street, N.W., Washington, D. C., and at the Moline Public Library, 504 - 17th Street, Moline, Illinois, for Quad Cities Unit No.1. A copy of items (2) and (3) t, may be obtained upon request addressed to the U. S. Nuclear Regulatory Commission, Washington, D. C. 20555, Attention: Director. Division of Licensing. ) Dated at Bethesda, Maryland, this 5th day of December,1980. FOR THE NUCLEAR REGULATORY COMMISSION - s Thomas A. Ippolito, Chief Operating Reactors Branch #2 ?? 'sion of Licensing ,}}